High Pressure Coolant Injection (HPCI) System Risk-Based Inspection Guide for Browns Ferry Nuclear Power Station

Energy Technology Data Exchange (ETDEWEB)

Wong, S.; DiBiasio, A.; Gunther, W. [Brookhaven National Lab., Upton, NY (United States)

1993-09-01

The High Pressure Coolant Injection (HPCI) system has been examined from a risk perspective. A System Risk-Based Inspection Guide (S-RIG) has been developed as an aid to HPCI system inspections at the Browns Ferry Nuclear Power Plant, Units 1, 2 and 3. The role of. the HPCI system in mitigating accidents is discussed in this S-RIG, along with insights on identified risk-based failure modes which could prevent proper operation of the system. The S-RIG provides a review of industry-wide operating experience, including plant-specific illustrative examples to augment the PRA and operational considerations in identifying a catalogue of basic PRA failure modes for the HPCI system. It is designed to be used as a reference for routine inspections, self-initiated safety system functional inspections (SSFIs), and the evaluation of risk significance of component failures at the nuclear power plant.

Pressure behavior in nuclear reactor containment following a loss of coolant accident

Energy Technology Data Exchange (ETDEWEB)

Khattab, M; Ibrahim, N A; Bedrose, C D [Reactors department, nuclear research center, atomic energy authority, Cairo, (Egypt)

1995-10-01

The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs.

Pressure behavior in nuclear reactor containment following a loss of coolant accident

International Nuclear Information System (INIS)

Khattab, M.; Ibrahim, N.A.; Bedrose, C.D.

1995-01-01

The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs

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

Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

International Nuclear Information System (INIS)

Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

2009-01-01

A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

High-pressure coolant effect on the surface integrity of machining titanium alloy Ti-6Al-4V: a review

Science.gov (United States)

Liu, Wentao; Liu, Zhanqiang

2018-03-01

Machinability improvement of titanium alloy Ti-6Al-4V is a challenging work in academic and industrial applications owing to its low thermal conductivity, low elasticity modulus and high chemical affinity at high temperatures. Surface integrity of titanium alloys Ti-6Al-4V is prominent in estimating the quality of machined components. The surface topography (surface defects and surface roughness) and the residual stress induced by machining Ti-6Al-4V occupy pivotal roles for the sustainability of Ti-6Al-4V components. High-pressure coolant (HPC) is a potential choice in meeting the requirements for the manufacture and application of Ti-6Al-4V. This paper reviews the progress towards the improvements of Ti-6Al4V surface integrity under HPC. Various researches of surface integrity characteristics have been reported. In particularly, surface roughness, surface defects, residual stress as well as work hardening are investigated in order to evaluate the machined surface qualities. Several coolant parameters (including coolant type, coolant pressure and the injection position) deserve investigating to provide the guidance for a satisfied machined surface. The review also provides a clear roadmap for applications of HPC in machining Ti-6Al4V. Experimental studies and analysis are reviewed to better understand the surface integrity under HPC machining process. A distinct discussion has been presented regarding the limitations and highlights of the prospective for machining Ti-6Al4V under HPC.

Real-time reactor coolant system pressure/temperature limit system

International Nuclear Information System (INIS)

Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

1991-01-01

This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

Nuclear reactor coolant and cover gas system

International Nuclear Information System (INIS)

George, J.A.; Redding, A.H.; Tower, S.N.

1976-01-01

A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

International Nuclear Information System (INIS)

Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

2003-03-01

Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). The present study focuses on detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model using RIA for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study in this year is to evaluate the success cri-teria of Aggressive Secondary Cooldown (ASC) in a Small Size Loss Of Coolant Accident (SBLOCA) without HPSI and to enhance the understanding of related thermal hydraulic behavior and phenomena. An effort was made to evaluate the system success criteria and a mission time for the recovery action by an operator to prevent the core damage for that accident scenario. The accident scenario for KSNP was a 2 inch coldleg break LOCA with a total loss of High Pressure Safety Injection (HPSI) and 1/2 Low Pressure Safety Injection (LPSI) available and perform-ing ASC limited by 55.6 .deg. C/hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip. It successively reached the LPSI condition for about 1.5hr after starting the ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria of 1204.4 .deg. C (2200 .deg. F). Sensitivity studies were performed for (1) cool-ant average temperature parameters, (2) ASC operation control method, (3) operation start time, (4) 1 inch break size. The present analysis identified thermal hydraulic phenomena and parameters affecting on the behavior, which consist of coolant break flow and inventory, parameters governing secondary heat removal, ASC operation control method, and its reference temperature parameters. In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that an operator should maintain the ade-quate ASC operation. However, it is necessary to evaluate the uncertainties arisen from the

Impact of high-pressure coolant supply on chip formation in milling

Science.gov (United States)

Klocke, F.; Döbbeler, B.; Lakner, T.

2017-10-01

Machining of titanium alloys is considered as difficult, because of their high temperature strength, low thermal conductivity and low E-modulus, which contributes to high mechanical loads and high temperatures in the contact zone between tool and workpiece. The generated heat in the cutting zone can be dissipated only in a low extent. When cutting steel materials, up to 75% of the process heat is transported away by the chips, contrary to only 25% when machining titanium alloys. As a result, the cutting tool heats up, which leads to high tool wear. Therefore, machining of titanium alloys is only possible with relatively low cutting speeds. This leads to low levels of productivity for milling processes with titanium alloys. One way to increase productivity is to use more cutting edges in tools with the same diameter. However, the limiting factor of adding more cutting edges to a milling tool is the minimum size of the chip spaces, which are sufficient for a stable chip evacuation. This paper presents experimental results on the chip formation and chip size influenced by high-pressure coolant supply, which can lead to smaller chips and to smaller sizes of the chip spaces, respectively. Both influences, the pressure of the supplied coolant and the volumetric flow rate were individually examined. Alpha-beta annealed titanium TiAl6V4 was examined in relation to the reference material quenched and tempered steel 42CrMo4+QT (AISI 4140+QT). The work shows that with proper chip control due to high-pressure coolant supply in milling, the number of cutting edges on the same diameter tool can be increased, which leads to improved productivity.

Pressure behaviour in a nuclear reactor containment following a loss of coolant accident

International Nuclear Information System (INIS)

KHattab, M.S.; Ibrahim, N.A.; Bedrose, S.D.

1994-01-01

The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break, is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The scenarios of small, medium and large LOCA at 2%, 15% and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The results of large LOCA showed good agreement with westinghouse calculations of the same design. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs., 1 tab

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.)

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

Theoretical study of hydraulic jump during circular horizontal hot leg injection in pressurized water reactor

International Nuclear Information System (INIS)

El Hawary, Shehab; Abu-Elyazeed, Osayed S.M.; Fahmy, Adel Alyan; Meglaa, Khairy

2016-01-01

Highlights: • The model is developed to predict the occurrence of onset hydraulic jump in a circular pipe. • Theoretical results are in agreement with experimental results and theory. • Effects of diameter of the injection pipe, Froude number and injected coolant mass are studied. - Abstract: One important phenomenon occurring during Loss of Coolant Accident (LOCA) is Counter-Current Flow Limitation (CCFL). The incidence of such CCFL is introduced by the onset of hydraulic jump. In the present work, a one dimensional model was modified to fit circular hot channel. The model was used to study the factors affecting the initial Froude number, the location of the occurrence of the hydraulic jump, and the critical coolant flow depth during circular horizontal hot leg injection in US-APWR Mitsubishi Reactor. The results showed good agreement with published experimental data of the Upper Plenum Test Facility (UPTF) at Mannheim, Germany. It was found that higher injected coolant mass flow rate increases the initial Froude number, the location of the occurrence of the hydraulic jump, and the critical injection depth divided by the diameter of the injection pipe. Such behavior is thought to be due to the increase of the inertia force by increasing of the injected coolant mass flow rate and the inverse of the diameter of the injection pipe. It was found also that, the location of the occurrence of hydraulic jump increases with decreasing load effect. Therefore, these results reveal that the avoidance of CCFL as well as hydraulic jump through hot leg at maximum load can be achieved by decreasing the distance between the injection point and the pressure vessel to below 0.3 m, and with diameter of 4 in (10.16 cm) as the design diameter of the injection pipe in US-APWR Mitsubishi Reactor. Moreover, the maximum critical depth (56 cm) is less than the diameter of the hot leg (78.74 cm) at an injected coolant mass flow of 400 kg/s, and with diameter of 4 in (10.16 cm) as the

Numerical study on coolant flow distribution at the core inlet for an integral pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Sun, Lin; Peng, Min Jun; Xia, Genglei; Lv, Xing; Li, Ren [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

2017-02-15

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

Transient performance analysis of pressurized safety injection tank with a partition

International Nuclear Information System (INIS)

Bae, Youngmin; Kim, Young In; Kim, Keung Koo

2015-01-01

Highlights: • Functional performance of safety injection tanks with a partition is evaluated. • Effects of key design parameters are scrutinized. • Distinctive features of the flow in multi-unit safety injection tanks are explored. - Abstract: A parametric study has been performed to evaluate the functional performance of a pressurized multi-unit safety injection tank, which would be considered as one of the candidates for a passive safety injection system in a nuclear power plant. The influences of key design parameters including the orifice size, initial gas fraction, and resistance coefficients and operating condition on the injection flow rate are scrutinized with a discussion of the relevant flow features such as the choked flow of gas through an orifice and two interconnected regions of differing gaseous pressure. The obtained results indicate that a multi-unit safety injection tank can passively control the injection flow rate and provide a stable safety injection over a relatively long period even in the case of drastic depressurization of a reactor coolant system

Interfacing systems loss of coolant accident (ISLOCA) pressure capacity methodology and Davis-Besse results

International Nuclear Information System (INIS)

Wesley, D.A.

1991-01-01

A loss of coolant accident resulting from the overpressurization by reactor coolant fluid of a system designed for low-pressure, low-temperature service has been identified as a potential contributor to nuclear power plant risk. In this paper, the methodology developed to assess the probability of failure as a function of internal pressure is presented, and sample results developed for the controlling failure modes and locations of four fluid systems at the Davis-Besse Plant are shown. Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The variability in the probability of failure is included, and the estimated leak rates or leak areas are given for the controlling modes of failure. For this evaluation, all failures are based on quasistatic pressures since the probability of dynamic effects resulting from such causes as water hammer have been initially judged to be negligible for the Davis-Besse plant ISLOCA

Steam generator for a pressurized-water coolant nuclear reactor

International Nuclear Information System (INIS)

Schroeder, H.J.; Berger, W.

1975-01-01

A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

Coolant clean-up and recycle systems

International Nuclear Information System (INIS)

Ito, Takao.

1979-01-01

Purpose: To increase the service life of mechanical seals in a shaft sealing device, eliminate leakages and improve the safety by providing a recycle pump for feeding coolants to a coolant clean-up device upon reactor shut-down and adapting the pump treat only low temperature and low pressure coolants. Constitution: The system is adapted to partially take out coolants from the pipeways of a recycling pump upon normal operation and feed them to a clean-up device. Upon reactor shut-down, the recycle pump is stopped and coolants are extracted by the recycle pump for shut-down into the clean-up device. Since the coolants are not fed to the clean-up device by the recycle pump during normal operation as conducted so far, high temperature and high pressure coolants are not directly fed to the recycle pump, thereby enabling to avoid mechanical problems in the pump. (Kamimura, M.)

Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

Energy Technology Data Exchange (ETDEWEB)

Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

1997-04-01

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

International Nuclear Information System (INIS)

Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

1997-01-01

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

Transient effects caused by pulsed gas and liquid injections into low pressure plasmas

International Nuclear Information System (INIS)

Ogawa, D; Goeckner, M; Overzet, L; Chung, C W

2010-01-01

The fast injection of liquid droplets into a glow discharge causes significant time variations in the pressure, the chemical composition of the gas and the phases present (liquid and/or solid along with gas). While the variations can be large and important, very few studies, especially kinetic studies, have been published. In this paper we examine the changes brought about in argon plasma by injecting Ar (gas), N 2 (gas) hexane (gas) and hexane (liquid droplets). The changes in the RF capacitively coupled power (forward and reflected), electron and ion density (n e , n i ), electron temperature (T e ) and optical emissions were monitored during the injections. It was found that the Ar injection (pressure change only) caused expected variations. The electron temperature reduced, the plasma density increased and the optical emission intensity remained nearly constant. The N 2 and hexane gas injections (chemical composition and pressure changes) also followed expected trends. The plasma densities increased and electron temperature decreased while the optical emissions changed from argon to the injected gas. These all serve to highlight the fact that the injection of evaporating hexane droplets in the plasma caused very little change. This is because the number of injected droplets is too small to noticeably affect the plasma, even though the shift in the chemical composition of the gas caused by evaporation from those same droplets can be very significant. The net conclusion is that using liquid droplets to inject precursors for low pressure plasmas is both feasible and controllable.

Analysis of loss-of-coolant accidents in pressurized water reactors

International Nuclear Information System (INIS)

Moldaschl, H.

1982-01-01

Analysis of loss-of-coolant accidents in pressurized water reactors -Quantification of the influence of leak size, control assembly worth, boron concentration and initial power by a dynamic operations criterion. Neutronic and thermohydraulic behaviour of a pressurized water reactor during a loss-of-coolant accident (LOCA) is mainly influenced by -change of fuel temperature, -void in the primary coolant. They cause a local stabilization of power density, that means that also in the case of small leaks local void is the main stabilization effect. As a consequence the increase of fuel temperature remains very small even under extremely hypothetical assumptions: small leak, positive reactivity feedback (positive coolant temperature coefficient, negative density coefficient) at the beginning of the accident and all control assemblies getting stuck. Restrictions which have been valid up to now for permitted start-up conditions to fulfill inherent safety requirements can be lossened substantially by a dynamic operations criterion. Burnable poisons for compensation of reactivity theorefore can be omitted. (orig.)

Low-pressure injection molding of alumina ceramics using a carnauba wax binder: preliminary results

Energy Technology Data Exchange (ETDEWEB)

Quevedo Nogueira, R.E.F.; Bezerra, A.C.; Santos, F.C. dos [Dept. de Engenharia Mecanica, Centro de Tecnologia-UFC, Fortaleza, CE (Brazil); Sousa, M.R. de; Acchar, W. [Dept. de Engenharia Mecanica, Univ. Federal do Rio Grande do Norte, UFRN-Campus Univ., Natal, RN (Brazil)

2001-07-01

Carnauba wax, a natural product from Northeastern Brazil, has found application in the processing of ceramics. However, the use of pure carnauba wax is not recommended due to its narrow melting range and poor mechanical properties. In the present work carnauba wax based organic vehicles with the addition of low-density polyethylene and stearic acid were developed for use in the low-pressure injection molding of alumina ceramics. Viscosimetric testing was employed for the determination of optimal composition of the organic vehicle. The optimal content of ceramic powder in the mixture was also determined. All the materials used are easily available in the Brazilian market. A simple ceramic part was injected at low pressures (0.6 MPa) using a semi-automatic injection molding machine. For this purpose a double cavity mold was designed and built. Preliminary results demonstrate the technical viability of the process using the organic vehicle developed. (orig.)

Pressurized-water coolant nuclear reactor steam generator

International Nuclear Information System (INIS)

Mayer, H.; Schroder, H.J.

1975-01-01

A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

Energy Technology Data Exchange (ETDEWEB)

Kryk, Holger, E-mail: h.kryk@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Hoffmann, Wolfgang [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany)

2014-12-15

Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

International Nuclear Information System (INIS)

Kryk, Holger; Hoffmann, Wolfgang; Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan

2014-01-01

Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products

Heat and fluid flow in accident of Fukushima Daiichi Nuclear Power Plant, Unit 3. Behaviour of high pressure coolant injection system (HPCI) based on thermodynamic model

International Nuclear Information System (INIS)

Maruyama, Shigenao

2014-01-01

In order to clarify the process of Accident of Fukushima Nuclear Plants, an accident scenario of Fukushima Daiichi Nuclear Power Plant, Unit 3 is analyzed from the data open to the public. Phase equilibrium process model was introduced in which the vapor and water are at saturation point in the vessels. The present accident scenario assumes that the high pressure coolant injection system (HPCI) did not worked properly, but the steam in the reactor pressure vessel (RPV) leaked through the turbine of HPCI to the suppression chamber since 12/3/2011 12:35. It is assumed that the Tsunami flooded the torus room where the suppression chamber was placed. Proposed accident scenario agrees with the data of the plant parameters obtained just after the accident. It is estimated that the water injection by HPIC was stopped since around at 13/3 19:00 and the water level in RPV decreased since then. It is estimated that the RPV broke at 14/3 8:55 and water could injected from fire engines due to the depression due to the rupture of RPV. There was little water left in RPV at the time of the rupture. If the present scenario is correct, the behavior that operators in the plant stopped HPCI at 13/3 2:42 did not affect seriously on the RPV rupture. If HPCI was working properly until the operators stopped it, the plant parameters obtained in the accident cannot be explained. (author)

Effect of cross-flow direction of coolant on film cooling effectiveness with one inlet and double outlet hole injection

Directory of Open Access Journals (Sweden)

Guangchao Li

2012-12-01

Full Text Available In order to study the effect of cross-flow directions of an internal coolant on film cooling performance, the discharge coefficients and film cooling effectiveness with one inlet and double outlet hole injections were simulated. The numerical results show that two different cross-flow directions of the coolant cause the same decrease in the discharge coefficients as that in the case of supplying coolant by a plenum. The different proportion of the mass flow out of the two outlets of the film hole results in different values of the film cooling effectiveness for three different cases of coolant supplies. The film cooling effectiveness is the highest for the case of supplying coolant by the plenum. At a lower blowing ratio of 1.0, the film cooling effectiveness with coolant injection from the right entrance of the passage is higher than that from the left entrance of the passage. At a higher blowing ratio of 2.0, the opposite result is found.

Coolant cleanup system for BWR type reactor

International Nuclear Information System (INIS)

Kinoshita, Shoichiro; Araki, Hidefumi.

1993-01-01

The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

Experiment data report for LOFT large-break loss-of-coolant experiment L2-5

International Nuclear Information System (INIS)

Bayless, P.D.; Divine, J.M.

1982-08-01

Selected pertinent and uninterpreted data from the third nuclear large break loss-of-coolant experiment (Experiment L2-5) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large [approx. 1000 MW(e)] commercial PWR operations. Experiment L2-5 simulated a double-ended offset shear of a cold leg in the primary coolant system. The primary coolant pumps were tripped within 1 s after the break initiation, simulating a loss of site power. Consistent with the loss of power, the starting of the high- and low-pressure injection systems was delayed. The peak fuel rod cladding temperature achieved was 1078 +- 13 K. The emergency core cooling system re-covered the core and quenched the cladding. No evidence of core damage was detected

Investigation of a hydrogen mitigation system during large break loss-of-coolant accident for a two-loop pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Dehjourian, Mehdi; Rahgoshay, Mohmmad; Jahanfamia, Gholamreza [Dept. of Nuclear Engineering, Science and Research Branch, Islamic Azad University of Tehran, Tehran (Iran, Islamic Republic of); Sayareh, Reza [Faculty of Electrical and Computer Engineering, Kerman Graduate University of Technology, Kerman (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

2016-10-15

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1979-08-01

The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

The effect of coolant quantity on local fuel–coolant interactions in a molten pool

International Nuclear Information System (INIS)

Cheng, Songbai; Matsuba, Ken-ichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Tohru; Tobita, Yoshiharu

2015-01-01

Highlights: • We investigate local fuel–coolant interactions in a molten pool. • As water volume increases, limited pressurization and mechanical energy observed. • Only a part of water is evaporated and responsible for the pressurization. - Abstract: Studies on local fuel–coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). Motivated by providing some evidence for understanding this interaction, in this study several experimental tests, with comparatively larger difference in coolant volumes, were conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are evaluated and compared. It is found that as water quantity increases, a limited pressure-buildup and the resultant mechanical energy release are observable. The performed analyses also suggest that only a part of water is probably vaporized during local FCIs and responsible for the pressurization and mechanical energy release, especially for those cases with much larger water volumes

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.

APPLICATION OF MULTIHOLE PRESSURE PROBE FOR RESEARCH OF COOLANT VELOCITY PROFILE IN NUCLEAR REACTOR FUEL ASSEMBLIES

Directory of Open Access Journals (Sweden)

S. M. Dmitriev

2015-01-01

Full Text Available Development of heat and mass transfer intensifiers is a major engineering task in the design of new and modernization of existing fuel assemblies. These devices create lateral mass flow of coolant. Design of intensifiers affects both the coolant mixing and the hydraulic resistance. The aim of this work is to develop a methodology of measuring coolant local velocity in the fuel assembly models with different mixing grids. To solve the problems was manufactured and calibrated multihole pressure probe. The air flow velocity measuring method with multihole pressure probe was used in the experimental studies on the coolant local hydrodynamics in fuel assemblies with mixing grids. Analysis of the coolant lateral velocity vector fields allowed to study the formation of the secondary vortex flows behind the mixing grids, and to determine the basic laws of coolant flow in experimental models. Quantitative data on the coolant flow velocity distribution obtained with a multihole pressure probe make possible to determine the magnitude of the flow lateral velocities in fuel rod gaps, as well as to determine the distance at which damping occurs during mixing. 

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.

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

Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

International Nuclear Information System (INIS)

Abass, N. M. N.

2012-02-01

A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

Development of a feed-and-bleed operation strategy with hybrid-SIT under low pressure condition of PWR

Energy Technology Data Exchange (ETDEWEB)

Jeon, In Seop, E-mail: jeoni@rpi.edu [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States); Han, Sang Hoon, E-mail: shhan2@kaeri.re.kr [Advanced Research Group, Korea Atomic Energy Research Institute, 70 Daedeok-daero 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Kang, Sang Hee, E-mail: sanghee.kang@khnp.co.kr [NSSS Design Group, Korea Hydro & Nuclear Power Co., Ltd., Central Research Institute, 70, 1312-beongil, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Kang, Hyun Gook, E-mail: hyungook@kaist.ac.kr [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States)

2017-04-01

Highlights: • The novel F&B operation strategy with H-SIT and LPSI is developed. • The effectiveness of the H-SITs is verified using thermo-hydraulic simulations. • Success criteria considered for the new F&B operation strategy is identified. • A PSA model of APR+ reflecting the new F&B strategy with H-SIT is developed. • A risk analysis of the proposed F&B operation strategy is performed. - Abstract: While safety functions in current nuclear power plants are mainly provided by active safety systems, recently passive safety systems are being combined with the active systems to strengthen accident mitigation capability and therefore enhance overall plant safety. To this end, securing long-term cooling of the core is of particular importance. This study considers the hybrid safety injection tank (H-SIT), a passive injection system, as a target component to develop a long-term cooling strategy using active and passive systems concurrently. In the feed-and-bleed (F&B) operation, one of the important long-term cooling strategies to maintain core safety in pressurized water reactors, low pressure safety injection (LPSI) pumps are typically considered inoperable as depressurization is first required, which leads to core dry-out before reaching LPSI operable pressure. This study investigates whether H-SITs, with the important design feature of passive coolant injection under any pressure condition of the primary coolant system, can make up the core during depressurization thereby allowing LPSI pumps to be used in F&B operation as an additional means of long-term cooling. The effectiveness of the H-SITs is verified using thermal-hydraulic simulations, and based on the results a novel F&B operation strategy with H-SITs and LPSI pumps is developed. A probabilistic safety assessment (PSA) model is then developed in order to assess the risk effect of the suggested strategy. PSA results demonstrate that the proposed strategy lowers core damage frequency in the target

Coupled analysis of passive safety injection and containment filtered venting for passive decay heat removal - 15140

International Nuclear Information System (INIS)

Kim, S.H.; Ham, J.H.; Jeong, Y.H.; Chang, S.H.

2015-01-01

Lots of interests for the safety of nuclear power plants have risen these days. The safety has to be continuously reviewed and enhanced in nuclear power plants currently operating as well as those designed and constructed in future. After the Fukushima accidents, many additional safety systems which can be applied to nuclear power plants in operation have been proposed. Those include alternating power source such as movable diesel generators and DC batteries in non-safety grade. Also, emergency preparedness for the prevention of a core damage accident was proposed to cope with the extended-SBO (station blackout) by using fire protection systems. In order to prevent the release of radioactive materials, safety systems for preserving the integrity of containment were proposed in two views of cooling and venting containment. Two approaches are effective for mitigating a severe accident. The design concept installing big water tanks besides containment at high level was proposed for various safety functions. One of the functions in the system is to inject the coolant from the elevated tank into a reactor vessel in the case of loss of coolant accident. When the pressure in reactor coolant system is sufficiently low, the coolant can be injected by gravity. If not, the depressurization in reactor vessel would be needed considering the containment pressure. Containment cooling in conventional pressurized water reactors is dependent on containment cooling pumps and sprays. Additional containment cooling systems cannot be simply and easily applied in the current nuclear power plants without major modifications. Therefore, for the operation of passive safety injection system, containment filtered venting system can be adopted for the depressurization of containment. In the design and operation of the passive safety injection system and the containment filtered venting system, main operating points related with open and close pressures in the filtered venting system were

The corrosion products in the coolant circuits of pressurized water nuclear power plants

International Nuclear Information System (INIS)

Darras, R.

1983-01-01

The characteristics of the corrosion products formed in the primary and secondary coolant circuits of light-water pressurized reactors are reviewed. The problem induced by the pollution of coolants and metallic surface are examined. Then, the recommendations to follow to minimize the disturbing effects of this pollution by the corrosion products are indicated [fr

Study on primary coolant system depressurization effect factor in pressurized water reactor

International Nuclear Information System (INIS)

Ji Duan; Cao Xuewu

2006-01-01

The progression of high-pressure core melting severe accident induced by very small break loss of coolant accident plus the loss of main feed water and auxiliary feed water failure is studied, and the entry condition and modes of primary cooling system depressurization during the severe accident are also estimated. The results show that the temperature below 650 degree C is preferable depressurization input temperature allowing recovery of core cooling, and the available and effective way to depressurize reactor cooling system and to arrest very small break loss of coolant accident sequences is activating pressurizer relief valves initially, then restoring the auxiliary feedwater and opening the steam generator relief valves. It can adequately reduce the primary pressure and keep the capacity loop of long-term core cooling. (authors)

On the transient pressure build-up in the full pressure safety shell of watercooled nuclear reactors after a loss of coolant accident

International Nuclear Information System (INIS)

Mansfeld, G.

1979-08-01

The thermo-and fluid-dynamic processes in a multichamber full pressure safety containment during a loss of coolant accident have been investigated. Comparison of the calculations carried out with the computer programs, in which ZOCO VI was used as being representative of similar programs, with the experimental results pointed out discrepancies in the determination of time dependent pressure, pressure difference and temperature curves. This led to the development of a new theoretical model and a program COFLOW which pays particular attention to the fluid dynamic processes in the initial phase of a loss of coolant accident. It can also be used to determine the maximum containment pressure towards the end of a loss of coolant accident. Comparison of the COFLOW results with experiments has shown that COFLOW provides a model and a procedure by which the physical processes in a multichamber full pressure safety containment can be simulated satisfactorily

Investigation of low pressure ES-SAGD

Energy Technology Data Exchange (ETDEWEB)

Ivory, J.; Zheng, R.; Nasr, T.; Deng, X.; Beaulieu, G.; Heck, G. [Alberta Research Council, Edmonton, AB (Canada)

2008-10-15

This paper described a scaled model experiment conducted to investigate the effectiveness of expanding solvent steam assisted gravity drainage (ES-SAGD) processes at low pressures. Lower SAGD pressures typically result in reduced oil production as a result of correspondingly lower steam temperatures. However, lower pressures may also result in a reduced steam to oil ratio (SOR) and a higher vaporization heat. Steam was injected into an injection well at 33 cm{sup 3} per minute and in a production well at 31 cm{sup 3} per minute. Steam and solvents were then co-injected into the injection well at a temperature of 206 degrees C. The experiment was history-matched and a parametric analysis was conducted using a simulation tool. The 2-D and 3-D field-scale simulations investigated the impact of operating pressures, injection rates; sub-cool; oil and gas phase diffusion and dispersion; live oil versus dead oil performance; and the use of drawdown when oil rates declined. Low pressure ES-SAGD was then compared with low-pressure SAGD. Results of the study suggested that production pressures, sub-cool and solvent concentrations are important parameters in ES-SAGD processes. At 1500 kPa production pressure and 10 degrees C sub-cool, the co-injection of solvent with steam increased average oil rates by 15 per cent more than the SAGD process. SOR was also reduced. 6 refs., 8 tabs., 20 figs.

Two-phase coolant pump model of pressurized light water nuclear reactors

International Nuclear Information System (INIS)

Santos, G.A. dos; Freitas, R.L.

1990-01-01

The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The homologous curves set up the complete performance of the pump and are input for accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

Cavitation and primary atomization in real injectors at low injection pressure condition

Science.gov (United States)

Dumouchel, Christophe; Leboucher, Nicolas; Lisiecki, Denis

2013-06-01

This experimental work investigates the influence of the geometry of GDI devices on primary atomization processes under low injection pressure and reduced back pressure. These pressure conditions ensure cavitating flows and observable atomization processes. Measurements include mass flux, structure velocity from high-speed visualizations and spray characterization with a laser diffraction technique. Super-cavitation regime and cavitation string, which have their own influence on the mass flux, develop independently in different injector regions. These regimes impact the flow pattern in the orifice and the subsequent atomization process. A possible interaction between cavitation string and super-cavitation is found to promote a hydraulic-flip-like regime and to deteriorate atomization quality. As far as the geometry of the injector is concerned, the profile of the orifice inlet and the roughness of the sac volume region are found to be important geometrical characteristics.

The origin and magnitude of pressures in fuel-coolant interactions

International Nuclear Information System (INIS)

Heer, W.; Jakeman, D.; Smith, B.L.

1987-01-01

A number of small scale experiments to simulate fuel coolant interaction (FCI) effects have been carried out using Freon and water. Contrary to the predictions of most current FCI models, only modest pressure transients are observed within the interaction region itself but large pressure spikes, near to or above critical Freon pressure, are seen at the boundaries of the region. Similar pressure amplification effects have been noticed in parallel experiments involving two phase mixtures. It is suggested that in both cases a water hammer type effect is the cause of the pressure spikes. These observations could form the basis of new thinking in FCI modelling. (author)

Pressurizing Behavior on Ingress of Coolant into Pebble Bed of Blanket of Fusion DEMO Reactor

International Nuclear Information System (INIS)

Daigo Tsuru; Mikio Enoeda; Masato Akiba

2006-01-01

Solid breeder blankets are being developed as candidate blankets for the Fusion DEMO reactor in Japan. JAEA is performing the development of the water cooled and helium cooled solid breeder blankets. The blanket utilizes ceramic breeder pebbles and multiplier pebbles beds cooled by high pressure water or high pressure helium in the cooling tubes placed in the blanket box structure. In the development of the blanket, it is very important to incorporate the safety technology as well as the performance improvement on tritium production and energy conversion. In the safety design and technology, coolant ingress in the blanket box structure is one of the most important events as the initiators. Especially the thermal hydraulics in the pebble bed in the case of the high pressure coolant ingress is very important to evaluate the pressure propagation and coolant flow behavior. This paper presents the preliminary results of the pressure loss characteristics by the coolant ingress in the pebble bed. Experiments have been performed by using alumina pebble bed (4 litter maximum volume of the pebble bed) and nitrogen gas to simulate the helium coolant ingress into breeder and multiplier pebble beds. Reservoir tank of 10 liter is filled with 1.0 MPa nitrogen. The nitrogen gas is released at the bottom part of the alumina pebble bed whose upper part is open to the atmosphere. The pressure change in the pebble bed is measured to identify the pressure loss. The measured values are compared with the predicted values by Ergun's equation, which is the correlation equation on pressure loss of the flow through porous medium. By the results of the experiments with no constraint on the alumina pebble bed, it was clarified that the measured value agreed in the lower flow rate. However, in the higher flow rate where the pressure loss is high, the measured value is about half of the predicted value. The differences between the measured values and the predicted values will be discussed from

Coolant mixing in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Hoehne, T; Grunwald, G

1998-10-01

The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

Compressed air injection technique to standardize block injection pressures.

Science.gov (United States)

Tsui, Ban C H; Li, Lisa X Y; Pillay, Jennifer J

2006-11-01

Presently, no standardized technique exists to monitor injection pressures during peripheral nerve blocks. Our objective was to determine if a compressed air injection technique, using an in vitro model based on Boyle's law and typical regional anesthesia equipment, could consistently maintain injection pressures below a 1293 mmHg level associated with clinically significant nerve injury. Injection pressures for 20 and 30 mL syringes with various needle sizes (18G, 20G, 21G, 22G, and 24G) were measured in a closed system. A set volume of air was aspirated into a saline-filled syringe and then compressed and maintained at various percentages while pressure was measured. The needle was inserted into the injection port of a pressure sensor, which had attached extension tubing with an injection plug clamped "off". Using linear regression with all data points, the pressure value and 99% confidence interval (CI) at 50% air compression was estimated. The linearity of Boyle's law was demonstrated with a high correlation, r = 0.99, and a slope of 0.984 (99% CI: 0.967-1.001). The net pressure generated at 50% compression was estimated as 744.8 mmHg, with the 99% CI between 729.6 and 760.0 mmHg. The various syringe/needle combinations had similar results. By creating and maintaining syringe air compression at 50% or less, injection pressures will be substantially below the 1293 mmHg threshold considered to be an associated risk factor for clinically significant nerve injury. This technique may allow simple, real-time and objective monitoring during local anesthetic injections while inherently reducing injection speed.

Upper plenum dump during reflood in PWR loss-of-coolant accident

International Nuclear Information System (INIS)

Sudo, Yukio; Griffith, Peter.

1981-01-01

Upper plenum dump during reflood in a large break loww-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood. The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnetude of water subcooling. (author)

Molding Properties of Inconel 718 Feedstocks Used in Low-Pressure Powder Injection Molding

Directory of Open Access Journals (Sweden)

Fouad Fareh

2016-01-01

Full Text Available The impact of binders and temperature on the rheological properties of feedstocks used in low-pressure powder injection molding was investigated. Experiments were conducted on different feedstock formulations obtained by mixing Inconel 718 powder with wax-based binder systems. The shear rate sensitivity index and the activation energy were used to study the degree of dependence of shear rate and temperature on the viscosity of the feedstocks. The injection performance of feedstocks was then evaluated using an analytical moldability model. The results indicated that the viscosity profiles of feedstocks depend significantly on the binder constituents, and the secondary binder constituents play an important role in the rheological behavior (pseudoplastic or near-Newtonian exhibited by the feedstock formulations. Viscosity values as low as 0.06 to 2.9 Pa·s were measured at high shear rates and high temperatures. The results indicate that a feedstock containing a surfactant agent exhibits the best moldability characteristics.

Mechanical energy yields and pressure volume and pressure time curves for whole core fuel-coolant interactions

Energy Technology Data Exchange (ETDEWEB)

Coddington, P [United Kingdom Atomic Energy Authority, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1979-10-15

In determining the damage consequences of a whole core Fuel-Coolant Interaction (FCI), one measure of the strength of a FCI that can be used and is independent of the system geometry is the constant volume mixing mechanical yield (often referred to as the Hicks-Menzies yield), which represents a near upper limit to the mechanical work of a FCI. This paper presents a recalculation of the Hicks-Menzies yields for UO{sub 2} and sodium for a range of initial fuel temperatures and fuel to coolant mass ratios, using recently published UO{sub 2} and sodium equation of state data. The work presented here takes a small number of postulated FCIs with as wide range as possible of thermal interaction parameters and determines their pressure-volume P(V) and pressure-time P(t) relations, using geometrical constraints representative of the reactor. Then by examining these P(V) and P(t) curves a representative pressure-relative volume curve or range of possible curves, for use in containment analysis, is recommended

RETRAN analysis of inter-system LOCA within the primary coolant pump

International Nuclear Information System (INIS)

Gangadharan, A.; Pratt, G.F.

1992-01-01

One example of an inter-system loss of coolant accident is the failure of the tubing within the primary coolant pump (PCP) thermal barrier heat exchanger. Such a failure would result in the entry of primary coolant into the component cooling water (CCW) system. The primary coolant flowrate through the break would rapidly pressurize the CCW system when the relief valves are too small. The piping in the CCW system at Palisades has a low pressure rating. Failures in this system outside the containment boundary could lead to primary coolant release to the atmosphere. RETRAN-02 was used to perform a simulation of the break in the PCP integral heat exchanger. The model included a detailed nodalization of the Byron-Jackson primary coolant pump internals leading up to the CCW system relief valves. Preliminary studies show the need for increased relief capacity in the CCW system. A case was run using a larger relief valve. Critical flow in the system upstream of the relief valves maintains the pressures in those volumes above the CCW design pressure. The pressures downstream from the relief valves and outside containment will be at or below the design pressure. This paper presents the results of the transient analysis

Fatigue management considering LWR coolant environments

International Nuclear Information System (INIS)

Park, Heung Bae; Jin, Tae eun

2000-01-01

Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)

Analysis of thermo-hydraulic behavior of coolant during discharge of pressurized high-temperature water

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Sobajima, Makoto; Sasaki, Shinobu; Onishi, Nobuaki; Shiba, Masayoshi

1978-01-01

The present report describes results of the analysis of the LOFT semiscale experiment No. 1011 using remodeled RELAP-3 code, performed at the Idaho National Engineering Laboratory to simulate a postulated loss-of-coolant accident in a pressurized water reactor. It was clarified through the analysis that coolant behavior during blowdown was influenced variously by the system components in the primary loop, comparing with coolant discharge from a pressure vessel. Good agreement was obtained between experimental and analytical results when phase separation was assumed in upper plenum and downcomer, since experimental data indicated existence of liquid level in those parts. It was also found that the use of the Wilson's equation to calculate bubble rise velocity and the use of discharge coefficient as the function of fluid quality at break location to calculate discharge flow rate resulted in good agreement with experimental data. (auth.)

N13 - based reactor coolant pressure boundary leakage system

International Nuclear Information System (INIS)

Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

1980-05-01

A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

Experimental Verification of Integrity of Low-Pressure Injection Piles Structure - Pile Internal Capacity

Science.gov (United States)

Pachla, Henryk

2017-12-01

The idea of strengthening the foundation using injection piles lies in transferring loads from the foundation to the piles anchorage in existing structure and formed in the soil. Such a system has to be able to transfer loads from the foundation to the pile and from the pile onto the soil. Pile structure often reinforced with steel element has to also be able to transfer such a loading. According to the rules of continuum mechanics, the bearing capacity of such a system and a deformation of its individual elements can be determined by way of an analysis of the contact problem of three interfaces. Each of these surfaces is determined by different couples of materials. Those surfaces create: pile-foundation anchorage, bonding between reinforcement and material from which the pile is formed and pilesoil interface. What is essential is that on the contact surfaces the deformation of materials which adhere to each other can vary and depends on the mechanical properties and geometry of these surfaces. Engineering practice and experimental research point out that the failure in such structures occurs at interfaces. The paper is concentrating on presenting the experiments on interaction between cement grout and various types of steel reinforcement. The tests were conducted on the special low pressure injection piles widely used to strengthen foundations of already existing structures of historical buildings due to the technology of formation and injection pressure.

A study on the effect of fluidic device installed in a safety injection tank on thermal-hydraulic phenomena of large break loss of coolant accident

International Nuclear Information System (INIS)

Chung, Young Jong; Bae, Kyoo Hwan; Song, Jin Ho; Sim, Suk Ku; Park, Jong Kyun

1999-03-01

The performance of the Safety Injection Tank (SIT) with fluidic device (advanced SIT) is analyzed for the large break loss of coolant accident (LBLOCA) using RELAP5/MOD3.1-KREM. First the case is analyzed using the conventional SIT. Among various cases the case with 4-split downcomer, discharge coefficient Cd=0.6, MCP trip with reactor trip and break location of cold leg discharge side with the pressurizer is found to be the most limiting case. For the same condition, the advanced SIT results the similar PCT, however it can maintain adequately the liquid level in the downcomer. By changing the ECCS location from the current injection to the cold leg elevations, PCT is improved by 75 K. (Author). 6 refs., 4 tabs., 54 figs

Pressure suppression device

International Nuclear Information System (INIS)

Yoshida, Toyokazu.

1976-01-01

Purpose: To provide a pressure suppression device for a gas cooled reactor wherein the coolant is discharged in a reactor building by a loss-of-coolant accident or the like, the increase in the pressure and temperature is controlled and thermal energy of the discharged coolant of high temperature and high pressure can be absorbed. Constitution: A low heat source unit is provided at the upper part in an inner space of a reactor building provided around the reactor, and at the upper part of the low heat source unit a stirring fan for mixing gas within the building, and a low heat source circulating the low heat source through a pipe is connected to the low heat source unit. The low heat source unit is provided with the pipe arranged in a spiral shape at the upper part of the space of the unit, and a large number of fins are provided at the outer surface of the pipe for increasing the transmission area and improve the heat exchange. When the coolant of high temperature and high pressure has been lost in the building, the thermal energy of the coolant is absorbed by the low heat source unit. (Aizawa, K.)

Safety assessment of the SMART design during SBLOCA tests using the high pressure safety injection pump of the SMART-ITL facility

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

SMART is a small-sized integral pressurized light water reactor designed by the Korea Atomic Energy Research Institute (KAERI) from 1997 and received standard design approval (SDA) by the Korean regulatory body in July 2012. Single reactor pressure vessel contains all of the main components including a pressurizer (PZR), steam generators (SG) and reactor coolant pumps (RCP) without any large-size pipes. Several tests to verify a safety and performance of SMART design were carried out. This paper introduces a comparison with three SBLOCA tests. Overall thermal-hydraulic phenomena were observed and showed a traditional trend to decrease a system pressure and temperature. A collapsed water level of the hot side indicated that the safety injection system was successfully operated to recover the reactor coolant system (RCS) and protect the core uncover. An SBLOCA test simulating a guillotine break on the SIS, SCS, and PSV was performed. It was enough to keep a steady-state condition before the SBLOCA test begins. An actuation signal as the boundary condition was properly simulated during the transient test. The scenarios of the SBLOCA in the SMART design were reproduced well using the SMART-ITL facility. The safety injection is effective to protect the core uncover as well as to cool down the RCS. All of the measured parameters show reasonable behaviors.

Safety assessment of the SMART design during SBLOCA tests using the high pressure safety injection pump of the SMART-ITL facility

International Nuclear Information System (INIS)

Bae, Hwang; Ryu, Sung Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yoon, Eun-Koo; Shin, Yong-Cheol; Min, Kyoung-Ho; Park, Jong-Kuk; Choi, Nam-Hyun; Bang, Yun-Gon; Seo, Chan-Jong; Yi, Sung-Jae; Park, Hyun-Sik

2016-01-01

SMART is a small-sized integral pressurized light water reactor designed by the Korea Atomic Energy Research Institute (KAERI) from 1997 and received standard design approval (SDA) by the Korean regulatory body in July 2012. Single reactor pressure vessel contains all of the main components including a pressurizer (PZR), steam generators (SG) and reactor coolant pumps (RCP) without any large-size pipes. Several tests to verify a safety and performance of SMART design were carried out. This paper introduces a comparison with three SBLOCA tests. Overall thermal-hydraulic phenomena were observed and showed a traditional trend to decrease a system pressure and temperature. A collapsed water level of the hot side indicated that the safety injection system was successfully operated to recover the reactor coolant system (RCS) and protect the core uncover. An SBLOCA test simulating a guillotine break on the SIS, SCS, and PSV was performed. It was enough to keep a steady-state condition before the SBLOCA test begins. An actuation signal as the boundary condition was properly simulated during the transient test. The scenarios of the SBLOCA in the SMART design were reproduced well using the SMART-ITL facility. The safety injection is effective to protect the core uncover as well as to cool down the RCS. All of the measured parameters show reasonable behaviors

Investigation of coolant mixture in pressurized water reactors at the Rossendorf mixing test facility ROCOM

International Nuclear Information System (INIS)

Grunwald, G.; Hoehne, T.; Prasser, H.M.; Richter, K.; Weiss, F.P.

1999-01-01

During the so-called boron dilution or cold water transients at pressurized water reactors too weakly borated water or too cold water, respectively, might enter the reactor core. This results in the insertion of positive reactivity and possibly leads to a power excursion. If the source of unborated or subcooled water is not located in all coolant loops but in selected ones only, the amount of reactivity insertion depends on the coolant mixing in the downcomer and lower plenum of the reactor pressure vessel (RPV). Such asymmetric disturbances of the coolant temperature or boron concentration might e.g. be the result of a failure of the chemical and volume control system (CVCS) or of a main steam line break (MSLB) that does only affect selected steam generators (SG). For the analysis of boron dilution or MSLB accidents coupled neutron kinetics/thermo-hydraulic system codes have been used. To take into account coolant mixing phenomena in these codes in a realistic manner, analytical mixing models might be included. These models must be simple and fast running on the one hand, but must well describe the real mixing conditions on the other hand. (orig.)

Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5

International Nuclear Information System (INIS)

Cardenas V, J.; Mugica R, C. A.; Lopez M, R.

2015-09-01

In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm 2 and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

A summary of the assessment of fuel behaviour, fission product release and pressure tube integrity following a postulated large loss-of-coolant accident

International Nuclear Information System (INIS)

Langman, V.J.; Weaver, K.R.

1984-05-01

The Ontario Hydro analyses of fuel and pressure tube temperatures, fuel behaviour, fission product release and pressure tube integrity for large break loss-of-coolant accidents in Bruce A or Pickering A have been critically reviewed. The determinations of maximum fuel temperatures and fission product release are very uncertain, and pressure tube integrity cannot be assured where low steam flows are predicted to persist for times on the order of minutes

Crack stability analysis of low alloy steel primary coolant pipe

Energy Technology Data Exchange (ETDEWEB)

Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

1997-04-01

At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

Experimental investigation of the effects of blowing conditions and Mach number on the unsteady behavior of coolant ejection through a trailing edge cutback

International Nuclear Information System (INIS)

Barigozzi, Giovanna; Armellini, Alessandro; Mucignat, Claudio; Casarsa, Luca

2012-01-01

Highlights: ► Flow visualization and PIV documented the presence of large coherent structures. ► The presence of coherent structures is documented up to the vane trailing edge. ► Shape and direction of rotation of vortices change with injection conditions. ► Vortices morphology influences the film cooling effectiveness distributions. ► A Mach number increase moves vortices closer to the wall. - Abstract: The present paper shows the results of an experimental investigation into the unsteadiness of coolant ejection at the trailing edge of a highly loaded nozzle vane cascade. The trailing edge cooling scheme features a pressure side cutback with film cooling slots, stiffened by evenly spaced ribs in an inline configuration. Cooling air is also ejected through two rows of cylindrical holes placed upstream of the cutback. Tests were performed with a low inlet turbulence intensity level (Tu 1 = 1.6%), changing the cascade operating conditions from low speed (M 2is = 0.2) up to high subsonic regime (M 2is = 0.6), and with coolant to main stream mass flow ratio varied within the 0.5–2.0% range. Particle Image Velocimetry (PIV) and flow visualizations were used to investigate the unsteady mixing process taking place between coolant and main flow downstream of the cutback, up to the trailing edge. For all the tested conditions, the results show the presence of large coherent structures, which presence is still evident up to the trailing edge. Their shape and direction of rotation change with injection conditions, as a function of coolant to mainstream velocity ratio, strongly influencing the thermal protection capability of the injected coolant flow. The Mach number increase is only responsible for a positioning of such vortical structures closer to the wall, while the Strouhal number almost remains unchanged.

Coolant leakage detection device

International Nuclear Information System (INIS)

Ito, Takao.

1983-01-01

Purpose: To surely detect the coolant leakage at a time when the leakage amount is still low in the intra-reactor inlet pipeway of FBR type reactor. Constitution: Outside of the intra-reactor inlet piping for introducing coolants at low temperature into a reactor core, an outer closure pipe is furnished. The upper end of the outer closure pipe opens above the liquid level of the coolants in the reactor, and a thermocouple is inserted to the opening of the upper end. In such a structure, if the coolants in the in-reactor piping should leak to the outer closure pipe, coolants over-flows from the opening thereof, at which the thermocouple detects the temperature of the coolants at a low temperature, thereby enabling to detect the leakage of the coolants at a time when it is still low. (Kamimura, M.)

Experimental Verification of Integrity of Low-Pressure Injection Piles Structure – Pile Internal Capacity

Directory of Open Access Journals (Sweden)

Pachla Henryk

2017-12-01

Full Text Available The idea of strengthening the foundation using injection piles lies in transferring loads from the foundation to the piles anchorage in existing structure and formed in the soil. Such a system has to be able to transfer loads from the foundation to the pile and from the pile onto the soil. Pile structure often reinforced with steel element has to also be able to transfer such a loading. According to the rules of continuum mechanics, the bearing capacity of such a system and a deformation of its individual elements can be determined by way of an analysis of the contact problem of three interfaces. Each of these surfaces is determined by different couples of materials. Those surfaces create: pile-foundation anchorage, bonding between reinforcement and material from which the pile is formed and pilesoil interface. What is essential is that on the contact surfaces the deformation of materials which adhere to each other can vary and depends on the mechanical properties and geometry of these surfaces. Engineering practice and experimental research point out that the failure in such structures occurs at interfaces. The paper is concentrating on presenting the experiments on interaction between cement grout and various types of steel reinforcement. The tests were conducted on the special low pressure injection piles widely used to strengthen foundations of already existing structures of historical buildings due to the technology of formation and injection pressure.

Coolant inlet device for nuclear reactors

International Nuclear Information System (INIS)

Ando, Hiroshi; Abe, Yasuhiro; Iwabuchi, Toshihiko; Yamamoto, Kenji.

1969-01-01

Herein disclosed is a coolant inlet device for liquid-metal cooled reactors which employs a coolant distributor serving also as a supporting means for the reactor core. The distributor is mounted within the reactor vessel so as to slide horizontally on supporting lugs, and is further slidably connected via a junction pipe to a coolant inlet conduit protruding through the floor of the vessel. The distributor is adapted to uniformly disperse the highly pressured coolant over the reactor core so as to reduce the stresses sustained by the reactor vessel as well as the supporting lugs. Moreover, the slidable nature of the distributor allows thermal shock and excessive coolant pressures to be prevented or alleviated, factors which posed major difficulties in conventional coolant inlet devices. (Owens, K. J.)

Research on coolant radiochemistry

International Nuclear Information System (INIS)

Yeon, Jei Won; Kim, W. H.; Park, Y. J.; Im, J. K.; Jung, Y. J.; Jee, K. Y.; Choi, K. C.

2004-04-01

The final objective of this study is to develop the technology on the reduction of radioactive material formed in reactor coolant circuit. The contents of this study are composed of the simulation of primary cooling system, chemistry measurement technology in the high-temperature high-pressure environments, and coolant chemistry control technology. The main results are as follows; High-temperature and high-pressure loop system was designed and fabricated, which is to inducing CRUD growth condition on the surface of cladding. The high-temperature pH measurement system was established with YSZ sensing electrode and Ag/AgCl reference electrode. The performance of pH electrode was confirmed in the temperature range 200∼280 .deg. C. Coolant chemistry control technologies such as the neutron irradiation technique of boric acid solution, the evaluation on high-temperature electrochemical behavior of coolant, and the measurement of physicochemical properties of micro-particles were developed. The results of this study can be useful for the understanding of chemical phenomena occurred in reactor coolant and for the study on the reduction of radioactive material in primary coolant, which will be carried out in the next research stage

Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

International Nuclear Information System (INIS)

Chopra, O.K.; Shack, W.J.

1998-03-01

The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented

Contact condensation effects in the main coolant pipe

International Nuclear Information System (INIS)

Haefner, W.; Fischer, K.

1990-01-01

Contact condensation effects may occur in a pressurized water reactor (PWR) after a loss of coolant accident (LOCA) when emergency core cooling (ECC) water is injected contact with escaping steam which is generated within the core. The condensation which takes place may cause a sudden depressurization leading to the formation of water slugs. The interaction between the transient condensation and the inertia of the flow may also result in large amplitude flow and pressure oscillations. These contact condensation effects are of great importance for the mass flow distribution and the coolant water supply to the reactor core. To examine those complex processes, large computer codes are necessary. The development and verification of analytical models requires greatly simplified flow boundary conditions from experiments and a sufficiently large base of experimental data. Separate models have been developed for interfacial exchange of mass, momentum and energy with respect to the associated flow regime. Therefore, an adequate description of the condensation process requires the modeling of two different topics: the prediction of the flow regime and the calculation of the interfacial exchange. (author)

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

Modeling the Rapid Boil-Off of a Cryogenic Liquid When Injected into a Low Pressure Cavity

Science.gov (United States)

Lira, Eric

2016-01-01

Many launch vehicle cryogenic applications require the modeling of injecting a cryogenic liquid into a low pressure cavity. The difficulty of such analyses lies in accurately predicting the heat transfer coefficient between the cold liquid and a warm wall in a low pressure environment. The heat transfer coefficient and the behavior of the liquid is highly dependent on the mass flow rate into the cavity, the cavity wall temperature and the cavity volume. Testing was performed to correlate the modeling performed using Thermal Desktop and Sinda Fluint Thermal and Fluids Analysis Software. This presentation shall describe a methodology to model the cryogenic process using Sinda Fluint, a description of the cryogenic test set up, a description of the test procedure and how the model was correlated to match the test results.

Simulation of small break loss of coolant accident using relap 5/ MOD 2 computer code

International Nuclear Information System (INIS)

Megahed, M.M.

1992-01-01

An assessment of relap 5 / MOD 2/Cycle 36.05 best estimate computer code capabilities in predicting the thermohydraulic response of a PWR following a small break loss of coolant accident is presented. The experimental data base for the evaluation is the results of Test S-N H-3 performed in the semi scale MOD-2 c Test facility which modeled a 0.5% small break loss of coolant accident with an accompanying failure of the high pressure injection emergency core cooling system. A conclusion was reached that the code is capable of making small break loss of coolant accident calculations efficiently. However, some of the small break loss of coolant accident related phenomena were not properly predicted by the code, suggesting a need for code improvement.9 fig., 3 tab

Determination of two dimensional axisymmetric finite element model for reactor coolant piping nozzles

International Nuclear Information System (INIS)

Choi, S. N.; Kim, H. N.; Jang, K. S.; Kim, H. J.

2000-01-01

The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively

Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

2016-05-15

KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

Research on Coolant Radiochemistry

International Nuclear Information System (INIS)

Ha, Yeong Keong; Kim, W. H.; Yeon, J. W.; Jung, Y. J.; Choi, K. C.; Choi, K. S.; Park, Y. J.; Cho, Y. H.

2007-06-01

The final objective of this study is to develop a method for reducing radioactive materials formed in the reactor coolant circuit. This second stage research was categorized into the following three subgroups: the development of the estimation technique of microscopic chemical variation at high temperatures and pressures, the fundamental study on the thermodynamics at high temperatures and pressures, and the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD. First, in the development of the estimation technique of microscopic chemical change at high temperatures and pressures, the technique for measuring coolant chemistry such as pH, conductivity and Eh was developed to be appropriate for the high temperature and pressure condition. The coolant chemistry measuring system including the self-devised high temperature pH sensor can be applied to the field of nuclear reactor and contribute on a large scale in the automation of the coolant chemistry control and the establishment of the real-time on-line measuring technique. Secondly, the dissociation constant of water and the solubility of metal oxides were measured in the fundamental study on the thermodynamics at high temperatures and pressures. Finally, in the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD, the careful investigation of the deposition phenomena of micro particles on the cladding surface showed that subcooled boiling and the dissolved hydrogen are the main factors responsible for the growth of CRUD. In addition, the basis was provided for the construction of a new particle behavior model in the reactor coolant circuit

Experiments on the behaviour of thermite melt injected into sodium: Final report on the THINA test results

International Nuclear Information System (INIS)

Huber, F.; Kaiser, A.; Peppler, W.

1994-01-01

During hypothetical accidents of fast breeder reactors the core melts and part of the core material inventory is ejected into the upper coolant plenum. As a consequence, a fuel to coolant thermal interaction occurs between the melt and the sodium. A series of simulating experiments was carried out in KfK/IRS to improve the knowledge about the phenomenology of molten fuel/coolant interactions and to support theoretical work on the safety of fast breeder reactors. In the tests, a thermite melt of up to 3270 K is injected from below into a sodium pool the temperature of which is between 770 and 820 K. The masses of the melt and the sodium are about five and 150 kg, respectively. Thermal interactions have been observed to occur as a sequence of small local pressure events mainly during the melt injection. Large-scale vapour explosions have not been observed. Generally, the conversion ratios of thermal to mechanical energy have been low. (author)

Coolant circuit water chemistry of the Paks Nuclear Power Plant

International Nuclear Information System (INIS)

Tilky, Peter; Doma, Arpad

1985-01-01

The numerous advantages of the proper selection of water chemistry parameters including low corrosion rate of the structural materials, hence the low-level activity build-up, depositions, radiation doses were emphasized. Major characteristics of water chemistry applied to the primary coolant of pressurized water reactors including neutral, slightly basic and strong basic ones are discussed. Boric acid is widely used to control reactivity. Primary coolant water chemistry of WWER type reactors which is based on the addition of ammonia and potassium hydroxide to boric acid is compared with that of other reactors. The demineralization of the total condensate of the steam turbines became a general trend in the water chemistry of the secondary coolant circuits. (V.N.)

Effects of molten material temperatures and coolant temperatures on vapor explosion

Institute of Scientific and Technical Information of China (English)

LI Tianshu; YANG Yanhua; YUAN Minghao; HU Zhihua

2007-01-01

An observable experiment facility for low-temperature molten materials to be dropped into water was set up in this study to investigate the mechanism of the vapor explosion. The effect of the fuel and coolant interaction(FCI) on the vapor explosion during the severe accidents of a fission nuclear reactor has been studied. The experiment results showed that the molten material temperature has an important effect on the vapor explosion behavior and pressure. The increase of the coolant temperature would decrease the pressure of the vapor explosion.

Numerical analysis of coolant mixing in the pressure vessel of WWER-440 type nuclear reactors

International Nuclear Information System (INIS)

Boros, I.; Aszodi, A.

2003-01-01

The precise description of the coolant mixing processes taking place in the reactor pressure vessel (RPV) of pressurized water nuclear reactors has an essential importance during power operation, as well as in case of incidental or accidental conditions. In this paper the detailed CFD model of the pressure vessel of a WWER-440 type reactor and calculations performed with this RPV model are presented. The CFD model of the pressure vessel contains all the important internal structural elements of the RPV. Sensitivity study on the effect of these elements was also carried out. Both steady-state and transient calculation were performed using the CFD code CFX-5.5.1. The results of the steady-state calculations give the so called mixing factors, i.e. the effect of each single primary loop at the core inlet. The mixing factors can be given for nominal circumstances (i.e. all main coolant pumps are working) or in case of less than six working MCPs. In order to validate the model the calculated mixing factors are compared with the values measured in the Paks NPP (Authors)

Coolant cleanup method in a nuclear reactor

International Nuclear Information System (INIS)

Kubota, Masayoshi; Nishimura, Shigeoki; Takahashi, Sankichi; Izumi, Kenkichi; Motojima, Kenji.

1983-01-01

Purpose : To effectively adsorb to remove low molecular weight organic substances from iron exchange resins for use in the removal of various radioactive nucleides contained in reactor coolants. Method : Reactor coolants are recycled by a main recyling pump in a nuclear reactor and a portion of the coolants is cooled and, thereafter, purified in a coolant desalter. While on the other hand, high pressure steams generated from the reactor are passed through a turbine, cooled in a condensator, eliminated with claddings or the likes by the passage through a filtration desalter using powderous ion exchange resins and then further passed through a desalter (filled with granular ion exchange resins). For instance, an adsorption and removing device for organic substances (resulted through the decomposition of ion exchange resins) precoated with activated carbon powder or filled with granular activated carbon is disposed at the downstream for each of the desalters. In this way, the organic substances in the coolants are eliminated to prevent the reduction in the desalting performance of the ion exchange resins caused by the formation of complexes between organic substances and cobalt in the coolants, etc. In this way, the coolant cleanup performance is increased and the amount of wasted ion exchange resins can be decreased. (Horiuchi, T.)

Physical characterisation of particles and rheological of a heterogeneous system used in low-pressure injection moulding

International Nuclear Information System (INIS)

Zampieron, Joao Vicente

2002-01-01

The powder injection moulding process is a recent technology, which offers as advantages a high production of complex geometry metal parts, with low cost, where secondary operations of machinery are unnecessary. The main of this thesis was centered on a coarse powders feedstock injection. The process begins with the composition of the mass, that is the combination of metal powders with organic binders. The following steps succeed injection in moulds, debinding, sintering and, if necessary, cleaning. For the formulation of the feedstock it is indispensable the characterisation of the powders. This is little mentioned in the open literature and brings up controversy among authors. At first, a series of powders characterisations of AISI 316 L stainless steel (below 25 μm) was adopted. The next step was to characterise the rheological behaviour of the feedstock using different rheological apparatus, so as to find the most appropriate equipment to the low-pressure powder injection molding process. The mass has to present a favourable rheological behaviour, which is low viscosity. The results of the physical characterisation were correlated among themselves and with the rheological characterisation. This was undertaken with the purpose of finding agreement among their values. Finally, the possibility of injection of water and gas atomised stainless steel coarse powders feedstock was studied. This presents as main advantage, the reduction of costs for the process. According to the literature, only powders with size below 25 μm are possible to be injected. Hence, starting from the physical characterisation of particles and rheological characterisation of the feedstock, the formulation of an appropriate mass was found for the coarse powders. These coarse powders were characterised by particles below 45 μm. In this case it was necessary to alter drastically the feedstock composition, using high amounts of wax, which lead to unstable rheological conditions. But, it was

Nonstationary pressure build up in full-pressure containments after a loss-of-coolant accident

International Nuclear Information System (INIS)

Mansfeld, G.

1977-01-01

The time histories of pressure, temperature and pressure difference during the pressure build up phase of a loss-of-coolant accident (LOCA) in the primary system in full-pressure containments of water cooled nuclear power reactors are treated. These are important for the design of such containments. The experiments within the German research program RS 50 ''Druckverteilung im Containment'' offered, for the first time, the opportunity to observe experimentally fluid-dynamic processes in a multiple divided full-pressure containment, and to test at the same time, computer codes which serve to describe the physical processes during the LOCA. The comparison of the results calculated by the computer codes ZOCO VI and DDIFF with the experimental results showed apparent deviations by special arrangements of the compartments and the vent flow paths of a model containment for the calculation of time dependent pressure-, temperature- and pressure difference-histories. The deviations lead to the development of the analytical model and computer code COFLOW. This new model was primarily designed to deal with the fluid-dynamic processes in the beginning phase of the blowdown as maximal pressure differences appear. Furthermore, it can be used to determine the maximum containment pressure, as well as for long term calculations. The analytical model and computer code COFLOW shows a better correlation between theory and experiment than previous codes

Reduced injection pressures using a compressed air injection technique (CAIT): an in vitro study.

Science.gov (United States)

Tsui, Ban C H; Knezevich, Mark P; Pillay, Jennifer J

2008-01-01

High injection pressures have been associated with intraneural injection and persistent neurological injury in animals. Our objective was to test whether a reported simple compressed air injection technique (CAIT) would limit the generation of injection pressures to below a suggested 1,034 mm Hg limit in an in vitro model. After ethics board approval, 30 consenting anesthesiologists injected saline into a semiclosed system. Injection pressures using 30 mL syringes connected to a 22 gauge needle and containing 20 mL of saline were measured for 60 seconds using: (1) a typical "syringe feel" method, and (2) CAIT, thereby drawing 10 mL of air above the saline and compressing this to 5 mL prior to and during injections. All anesthesiologists performed the syringe feel method before introduction and demonstration of CAIT. Using CAIT, no anesthesiologist generated pressures above 1,034 mm Hg, while 29 of 30 produced pressures above this limit at some time using the syringe feel method. The mean pressure using CAIT was lower (636 +/- 71 vs. 1378 +/- 194 mm Hg, P = .025), and the syringe feel method resulted in higher peak pressures (1,875 +/- 206 vs. 715 +/- 104 mm Hg, P = .000). This study demonstrated that CAIT can effectively keep injection pressures under 1,034 mm Hg in this in vitro model. Animal and clinical studies will be needed to determine whether CAIT will allow objective, real-time pressure monitoring. If high pressure injections are proven to contribute to nerve injury in humans, this technique may have the potential to improve the safety of peripheral nerve blocks.

Analysis Of Primary Coolant Suction Side Pressure In The Delay Chamber Of The RSG-GAS

International Nuclear Information System (INIS)

Dibyo, Sukmanto

2000-01-01

Delay chamber is a tank to delay flow that located in the primary cooling suction side of RSG-GAS. A void occurred when operation reactor caused by too high the delta P at inlet suction pump. The condition may be avoided by using one line mode of the cooling flow. The analysis show that void volume in the delay chamber is occurred because the coolant negative pressure lowers the saturation pressure should be avoided though decreasing the delta P until about 0.1 bar at about 45 exp 0 C. Solution suggested are to use bypass flow from the spent fuel to the delay chamber. Coolant temperature can be also decreased by decreasing the power level of the reactor as well as improving the heat exchanger and cooling tower performances

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

Energy Technology Data Exchange (ETDEWEB)

Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

2017-08-15

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

Analysis on transient hydrodynamic characteristics of cavitation process for reactor coolant pump

International Nuclear Information System (INIS)

Wang Xiuli; Wang Peng; Yuan Shouqi; Zhu Rongsheng; Fu Qiang

2014-01-01

The reactor coolant pump hydrodynamic characteristics at different cavitation conditions were studied by using flow field analysis software ANSYS CFX, and the corresponding data were processed and analyzed by using Morlet wavelet transform and fast Fourier transform. The results show that gas content presents the law of exponential function with the pressure reduction or time increase. In the cavitation primary condition, the pulsation frequency of head for the reactor coolant pump is mainly low frequency, and the main frequency of pressure pulsation is still rotation frequency while the effect of the pressure pulsation caused by cavitation on main frequency is not obvious. With the development of cavitation, the pressure fluctuation induced by cavitation becomes more serious especially for the main frequency, secondary frequency and pulsating amplitude while the head pulsation frequency is given priority to low frequency pulse. Under serious cavitation condition, the head pulsation frequency is given priority to irregular changes of pulse high frequency, and also contains almost regular changes of low frequency. (authors)

Integrated main coolant pumps for pressurized-water reactors

International Nuclear Information System (INIS)

Wieser, R.

1975-01-01

The efficiency of an integrated main coolant pump for PWR's is increased. For this purpose, the pump is installed eccentric relative to the vertical axis of the U-type steam generator in the three-section HP chamber in such a way that its impeller wheel and the shell of the latter penetrate into the outlet chamber. The axis of the pump lies in the vertical plane of symmetry of the outlet chamber of the steam generator. The suction tube is arranged in the outlet chamber. To allow it to be installed, it is manufactured out of several parts. The diffusor tube, which is also made of several components, is attached to the horizontal separation plate between the outlet chamber and the pressure chamber so as to penetrate into it. To improve the outflow conditions at the diffusor tube, a plowshare-shaped baffle shield is installed between the diffusor tube and the HP chamber. Moreover, in order to improve the outflow conditions from the pump and from the pressure chamber, the outflow opening of the pressure chamber is put into the cylindrical shell of the HP chamber. In this way, the tensioning anchor is located between the pump and the outlet opening. (DG/RF) [de

The choice between two designs for the safety-injection system of a pressurized-water reactor, using probabilistic methods

International Nuclear Information System (INIS)

Villemeur, Alain

1982-01-01

A probabilistic study has been carried out to compare two designs for the safety-injection circuit of a pressurized-water reactor. It appears that unavailability of the circuit after an accident involving loss of coolant decreases little when one moves from a 2-line to a 3-line system. These results are compared with the disadvantages arising from increased redundancy, and in particular the increased cost of the installations. The 2-line circuit appears the optimum one on the basis of cost and reliability criteria. It has been chosen for the 1300-MWe units [fr

Simulation of coolant mixing in pressure vessel reactors

International Nuclear Information System (INIS)

Hoehne, T.

2003-06-01

The work was aimed at the experimental investigation and numerical simulation of coolant mixing in the downcomer and the lower plenum of PWRs. Generally, the coolant mixing is of relevance for two classes of accident scenarios - boron dilution and cold water transients. For the investigation of the relevant mixing phenomena, the Rossendorf test facility ROCOM has been designed. ROCOM is a 1:5 scaled Plexiglas trademark model of the PWR Konvoi allowing conductivity measurements by wire mesh sensors and velocity measurements by the LDA technique. The CFD calculations were carried out with the CFD-code CFX-4. For the design of the facility, calculations were performed to analyze the scaling of the model. It was found, that the scaling of 1:5 to the prototype meets both: physical and economical demands. Flow measurements and the corresponding CFD calculations in the ROCOM downcomer under steady state conditions showed a Re number independency at nominal flow rates. The flow field is dominated by recirculation areas below the inlet nozzles. Transient flow measurements with high performance LDA-technique showed in agreement with CFX-4 results, that in the case of the start up of a pump after a laminar stage large vortices dominate the flow. In the case of stationary mixing, the maximum value of the averaged mixing scalar at the core inlet was found in the sector below the inlet nozzle, where the tracer was injected. At the start-up case of one pump due to a strong impulse driven flow at the inlet nozzle the horizontal part of the flow dominates in the downcomer. The injection is distributed into two main jets, the maximum of the tracer concentration at the core inlet appears at the opposite part of the loop where the tracer was injected. Additionally, the stationary three-dimensional flow distribution in the downcomer and the lower plenum of a VVER-440/V-230 reactor was calculated with CFX-4. The comparison with experimental data and an analytical mixing model showed a

Thrust Vectoring of a Continuous Rotating Detonation Engine by Changing the Local Injection Pressure

International Nuclear Information System (INIS)

Liu Shi-Jie; Lin Zhi-Yong; Sun Ming-Bo; Liu Wei-Dong

2011-01-01

The thrust vectoring ability of a continuous rotating detonation engine is numerically investigated, which is realized via increasing local injection stagnation pressure of half of the simulation domain compared to the other half. Under the homogeneous injection condition, both the flow-field structure and the detonation wave propagation process are analyzed. Due to the same injection condition along the inlet boundary, the outlines of fresh gas zones at different moments are similar to each other. The main flow-field features under thrust vectoring cases are similar to that under the baseline condition. However, due to the heterogeneous injection system, both the height of the fresh gas zone and the pressure value of the fresh gas in the high injection pressure zone are larger than that in the low injection pressure zone. Thus the average pressure in half of the engine is larger than that in the other half and the thrust vectoring adjustment is realized. (fundamental areas of phenomenology(including applications))

Dryout heat flux in a debris bed with forced coolant flow from below

International Nuclear Information System (INIS)

Bang, Kwang-Hyun; Kim, Jong-Myung

2004-01-01

The objective of the present study is to experimentally investigate the enhancement of dryout heat flux in debris beds with coolant flow from below. The experimental facility consists mainly of an induction heater (40 kW, 35 kHz), a double-wall quartz-tube test section containing steel-particle bed and coolant injection and recovery condensing loop. A fairly uniform heating of particle bed was achieved by induction heating. This paper reports the experimental data for 5 mm particle bed and 300 mm bed height. The dryout heat rate data were obtained of both top-flooding case and forced coolant injection from below with the injection mass flux up to 1.5 kg/m 2 s. For the top-flooded case, the volumetric dryout heat rate was about 4 MW/m 3 and it increased as the rate of coolant injection from below was increased. At the coolant injection mass flux of 1.5 kg/m 2 s, the volumetric dryout heat rate was about 10 MW/m 3 , the enhancement factor was more than two. (author)

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

International Nuclear Information System (INIS)

Jones, R.D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

Science.gov (United States)

Jones, Robert D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

Development of in-situ laser cutting technique for removal of single selected coolant channel from pressurized heavy water reactor

International Nuclear Information System (INIS)

Vishwakarma, S.C.; Upadhyaya, B.N.

2016-01-01

We report on the development of a pulsed Nd:YAG laser based cutting technique for removal of single coolant channel from pressurized heavy water reactor (PHWR). It includes development of special tools/manipulators and optimization of laser cutting process parameters for cutting of liner tube, end fitting, bellow lip weld joint, and pressure tube stubs. For each cutting operation, a special tool with precision motion control is utilized. These manipulators/tools hold and move the laser cutting nozzle in the required manner and are fixed on the same coolant channel, which has to be removed. This laser cutting technique has been successfully deployed for removal of selected coolant channels Q-16, Q-15 and N-6 of KAPS-2 reactor with minimum radiation dose consumption and in short time. (author)

Infusion pressure and pain during microneedle injection into skin of human subjects

Science.gov (United States)

Gupta, Jyoti; Park, Sohyun; Bondy, Brian; Felner, Eric I.; Prausnitz, Mark R.

2011-01-01

Infusion into skin using hollow microneedles offers an attractive alternative to hypodermic needle injections. However, the fluid mechanics and pain associated with injection into skin using a microneedle have not been studied in detail before. Here, we report on the effect of microneedle insertion depth into skin, partial needle retraction, fluid infusion flow rate and the co-administration of hyaluronidase on infusion pressure during microneedle-based saline infusion, as well as on associated pain in human subjects. Infusion of up to a few hundred microliters of fluid required pressures of a few hundred mmHg, caused little to no pain, and showed weak dependence on infusion parameters. Infusion of larger volumes up to 1 mL required pressures up to a few thousand mmHg, but still usually caused little pain. In general, injection of larger volumes of fluid required larger pressures and application of larger pressures cause more pain, although other experimental parameters also played a significant role. Among the intradermal microneedle groups, microneedle length had little effect; microneedle retraction lowered infusion pressure but increased pain; lower flow rate reduced infusion pressure and kept pain low; and use of hyaluronidase also lowered infusion pressure and kept pain low. We conclude that microneedles offer a simple method to infuse fluid into the skin that can be carried out with little to no pain. PMID:21684001

Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment

International Nuclear Information System (INIS)

El-Genk, M.S.; Hobbins, R.R.

1981-01-01

The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO 2 fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm 3 /s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO 2 fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

Directory of Open Access Journals (Sweden)

Hwang Bae

2017-08-01

Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

International Nuclear Information System (INIS)

Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

2000-01-01

A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

Influence of fuel injection pressures on Calophyllum inophyllum methyl ester fuelled direct injection diesel engine

International Nuclear Information System (INIS)

Nanthagopal, K.; Ashok, B.; Karuppa Raj, R. Thundil

2016-01-01

Highlights: • Effect of injection pressure of Calophyllum inophyllum biodiesel is investigated. • Engine characteristics of 100% Calophyllum inophyllum biodiesel has been performed. • Calophyllum inophyllum is a non-edible source for biodiesel production. • Increase in injection pressure of biodiesel, improves the fuel economy. • Incylinder pressure characteristics of biodiesel follows similar trend as of diesel. - Abstract: The trend of using biodiesels in compression ignition engines have been the focus in recent decades due to the promising environmental factors and depletion of fossil fuel reserves. This work presents the effect of Calophyllum inophyllum methyl ester on diesel engine performance, emission and combustion characteristics at different injection pressures. Experimental investigations with varying injection pressures of 200 bar, 220 bar and 240 bar have been carried out to analyse the parameters like brake thermal efficiency, specific fuel consumption, heat release rate and engine emissions of direct injection diesel engine fuelled with 100% biodiesel and compared with neat diesel. The experimental results revealed that brake specific fuel consumption of C. inophyllum methyl ester fuelled engine has been reduced to a great extent with higher injection pressure. Significant reduction in emissions of unburnt hydrocarbons, carbon monoxide and smoke opacity have been observed during fuel injection of biodiesel at 220 bar compared to other fuel injection pressures. However oxides of nitrogen increased with increase in injection pressures of C. inophyllum methyl ester and are always higher than that of neat diesel. In addition the combustion characteristics of biodiesel at all injection pressures followed a similar trend to that of conventional diesel.

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.)

Device for preventing coolant outflow in a reactor

International Nuclear Information System (INIS)

Nemoto, Kiyomitsu; Mochizuki, Keiichi.

1975-01-01

Object: To prevent outflow of coolant from a reactor vessel even in an occurrence of leaking trouble at a low position in a primary cooling system or the like in the reactor vessel. Structure: An inlet at the foremost end of a coolant inlet pipe inserted into a reactor vessel is arranged at a level lower than a core, and a check valve is positioned at a level higher than the core in a rising portion of the inlet. In normal condition, the check valve is pushed up by discharge pressure of a main circulating pump and remains closed, and hence, producing no flow loss of coolant, sodium. However, when a trouble such as rupture occurs at the lower position in the primary cooling system, the attractive force for allowing the coolant to back-flow outside the reactor vessel and the load force of the coolant within the reactor vessel cause the check valve to actuate, as a consequence of which a liquid level of the coolant downwardly moves to the position of the check valve to intake the cover gases into a gas intake, thereby cutting off a flow passage of the coolant to stop outflow thereof. (Kamimura, M.)

Analytical prediction on the pump-induced pulsating pressure in a reactor coolant pipe

International Nuclear Information System (INIS)

Lee, K.B.; Im, I.Y.; Lee, S.K.

1992-01-01

An analytical method is presented for predicting the amplitudes of pump-induced fluctuating pressures in a reactor coolant pipe using a linear transformation technique which reduces a homogeneous differential equation with non-homogeneous boundary conditions into a nonhomogeneous differential equation with homogeneous boundary conditions. At the end of the pipe, three types of boundary conditions are considered-open, closed and piston-spring supported. Numerical examples are given for a typical reactor. Comparisons of measured pressure amplitudes in the pipe with model prediction are shown to be in good agreement for the forcing frequencies. (author)

Analysis of containment pressure and temperature changes following loss of coolant accident (LOCA)

International Nuclear Information System (INIS)

Nguyen Van Thai; Kieu Ngoc Dung

2015-01-01

This paper present a preliminary thermal-hydraulics analysis of AP1000 containment following loss of coolant accident events such as double-end cold line break (DECLB) or main steam line break (MSLB) using MELCOR code. A break of this type will produce a rapid depressurization of the reactor pressure vessel (primary system) and release initially high pressure water into the containment followed by a much smaller release of highly superheated steam. The high pressure liquid water will flash and rapidly pressurize the containment building. The performance of passive containment cooling system for steam removal by condensation on large steel containment structure is a major contributing process, controlling the pressure and temperature maximum reached during the accident event. The results are analyzed, discussed and compared with the similar work done by Sandia National Laboratories. (author)

Cavity Pressure Behaviour in Micro Injection Moulding

DEFF Research Database (Denmark)

Griffiths, C.A.; Dimov, S.S.; Scholz, S.

2010-01-01

as well as with the filling of the cavity by the polymer melt. In this paper, two parameters derived from cavity pressure over time (i.e. pressure work). The influence of four µIM parameters (melt temperature, mould temperature, injection speed, aand packing pressure) on the two pressure-related outputs...... has been investigated by moulding a micro fluidic component on three different polymers (PP, ABS, PC) using the design of experiment approach. Similar trends such as the effects of a higher injection speed in decreasing the pressure work and of a lower temperature in decreasing pressure rate have been......Process monitoring of micro injection moulding (µIM) is of crusial importance to analyse the effect of different parameter settings on the process and to assess its quality. Quality factors related to cavity pressure can provide useful information directly connected with the dyanmics of the process...

Partial Discharge Measurements in HV Rotating Machines in Dependence on Pressure of Coolant

Directory of Open Access Journals (Sweden)

I. Kršňák

2002-01-01

Full Text Available The influence of the pressure of the coolant used in high voltage rotating machines on partial discharges occurring in stator insulation is discussed in this paper. The first part deals with a theoretical analysis of the topic. The second part deals with the results obtained on a real generator in industrial conditions. Finally, theoretical assumptions and obtained results are compared.

Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

Energy Technology Data Exchange (ETDEWEB)

Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

Energy Technology Data Exchange (ETDEWEB)

Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

International Nuclear Information System (INIS)

Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun; Park, Rae Joon; Kim, Sang Baik

1997-01-01

This paper presents results of experimental studies on the heat transfer and solidifcation of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. As a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 .deg. C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleight number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer

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...

Primary coolant feed and bleed operating regions for the Midland Plant

International Nuclear Information System (INIS)

Tsai, M.S.

1985-01-01

Operating regions for primary coolant feed and bleed cooling are developed for the Midland Plant using core decay heat, the high-pressure injection (HPI) system capacity, and flow rate relief through the power-operated relief valve (PORV). This mode of cooling is used for accident scenarios in which the normal core cooling means of a nuclear power plant is lost because of loss of water inventory in the steam generators. The HPI flow is based on the capacities of one and two pumps. Saturated steam, saturated water, and subcooled water are considered to be possible states of the fluid being relieved through the PORV. In estimating the PORV relief rate, flow equations are derived from the Electric Power Research Institute test data obtained from the same model and size valve that is used in the Midland Plant. For easy reference by operators, the operating region is displayed on a plane of reactor coolant system pressure and temperature. The technique developed for the Midland Plant provides a convenient method for examining the feed and bleed cooling capability for a nuclear power plant that employs a pressurized water reactor system

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

Reactor coolant pumps for nuclear reactors

International Nuclear Information System (INIS)

Harand, E.; Richter, G.; Tschoepel, G.

1975-01-01

A brake for the pump rotor of a main coolant pump or a shutoff member on the pump are provided in order to prevent excess speeds of the pump rotor. Such excess speeds may occur in PWR type reactors with water at a pressure below, e.g., 150 bars if there is leakage from a coolant line associated with the main coolant pump. As a brake, a centrifugal brake depending upon the pump speed or a brake ring arranged on the pump housing and acting on the pump rotor, which ring would be activated by pressure differentials in the pump, may be used. If the pressure differences between suction and pressure sockets are very small, a controlled hydraulic increase of the pressure force on the brake may also be provided. Furthermore, a turbine brake may be provided. A slide which is automatically movable in closing position along the pump rotor axis is used as a shutoff element. It is of cylindrical configuration and is arranged concentrically with the rotor axis. (DG) [de

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

Flow boiling test of GDP replacement coolants

International Nuclear Information System (INIS)

Park, S.H.

1995-01-01

The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C 4 F 10 and C 4 F 8 , were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C 4 F 10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C 4 F 10 weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd

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.

Microbially Enhanced Oil Recovery by Sequential Injection of Light Hydrocarbon and Nitrate in Low- And High-Pressure Bioreactors.

Science.gov (United States)

Gassara, Fatma; Suri, Navreet; Stanislav, Paul; Voordouw, Gerrit

2015-10-20

Microbially enhanced oil recovery (MEOR) often involves injection of aqueous molasses and nitrate to stimulate resident or introduced bacteria. Use of light oil components like toluene, as electron donor for nitrate-reducing bacteria (NRB), offers advantages but at 1-2 mM toluene is limiting in many heavy oils. Because addition of toluene to the oil increased reduction of nitrate by NRB, we propose an MEOR technology, in which water amended with light hydrocarbon below the solubility limit (5.6 mM for toluene) is injected to improve the nitrate reduction capacity of the oil along the water flow path, followed by injection of nitrate, other nutrients (e.g., phosphate) and a consortium of NRB, if necessary. Hydrocarbon- and nitrate-mediated MEOR was tested in low- and high-pressure, water-wet sandpack bioreactors with 0.5 pore volumes of residual oil in place (ROIP). Compared to control bioreactors, those with 11-12 mM of toluene in the oil (gained by direct addition or by aqueous injection) and 80 mM of nitrate in the aqueous phase produced 16.5 ± 4.4% of additional ROIP (N = 10). Because toluene is a cheap commodity chemical, HN-MEOR has the potential to be a cost-effective method for additional oil production even in the current low oil price environment.

Cold leg injection reflood test results in the SCTF Core-I under constant system pressure

International Nuclear Information System (INIS)

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

1990-08-01

The Slab Core Test Facility (SCTF) was constructed to investigate two-dimensional thermal-hydrodynamics in the core and the interaction in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a postulated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The present report describes the analytical results on the system behavior observed in the SCTF Core-I cold leg injection tests, S1-14 (Run 520), S1-15 (521), S1-16 (522), S1-17 (523), S1-20 (530), S1-21 (531), S1-23 (536) and S1-24 (537), performed under constant system pressure condition during transient. Major discussion items are: (1) steam binding, (2) U-tube oscillations, (3) bypass of ECC water (4) core cooling behavior, (5) effect of vent valve and (6) other parameter effects. These results give us very useful information and suggestion on reflood behavior. (author)

An experimental study on the effects of high-pressure and multiple injection strategies on DI diesel engine emissions

KAUST Repository

Yang, Seung Yeon

2013-03-25

An experimental study on effects of high-pressure injections in conjunction with split fuel injections were conducted on an AVL single cylinder DI diesel engine. Various injection schemes were studied through the use of an electronically controlled, common rail injection system capable of injection pressures up to 200 MPa and a maximum of six injections per combustion event. Up to 100 MPa of the fuel injection pressure, the higher injection pressures create faster combustion rates that result in the higher in-cylinder gas temperatures as compared to conventional low-pressure fuel injection systems. When applying high-pressure injections, particulate emission reductions of up to 50% were observed with no change in hydrocarbon emissions, reductions of CO emissions and only slightly higher NOx emissions. Over 100 MPa, on the other hand, the higher injection pressures still reduced up to almost zero-level of particulate emission, at the same time that the NO emission is reduced greatly. Under these high-pressure injection conditions, strong correlations between soot and CO emissions were observed, which compete for the oxidizing OH species. Multiple or split high-pressure injections also investigated as a means to decrease particulate emissions. As a result, a four-split injection strategy resulted in a 55% reduction in particulates and with little or no penalty on NOx emissions. The high pressure split injection strategy with EGR was more effective in reducing particulate and CO emissions simultaneously. Copyright © 2013 SAE International and Copyright © 2013 TSAE.

''Iodine delivery rate'' with catheterangiography under pressure conditions of hand injection

International Nuclear Information System (INIS)

Busch, H.P.; Stocker, K.P.

1998-01-01

Purpose: The aim of this study was to record the flow-rate and to calculate the 'iodine delivery rate' (IDR) of contrast media of various viscosities when the contrast media are injected by hand. Methods: Five different catheters for coronary angiography were tested with the injection system Medral Mark V Plus. Injections were performed with pressures of 100, 200 and 400 PSI. The contrast media examined were Imeron 350, Imeron 400, Omnipaque 350 and Ultravist 370. The IDR was calculated on the basis of the measured flow rate and the Iodine content of the contrast medium. Results: As was expected, the IDR was higher as the pressure increased. In addition to the iodine content the viscosity of the contrast medium is a very important factor for the IDR. At both 100 PSI and 200 PSI the increase of the IDR was higher with Imeron 350 than with Imeron 400. The comparison of contrast media with identical iodine contents but differing viscosities (Imeron 350, Omnipaque 350) clearly showed that the IDR depended on viscosity. Conclusion: The 'iodine delivery rate' is a decisive factor in the opacification of arterial vessels. Both iodine content and viscosity of a contrast medium are important for the IDR. Because of the low pressure at manual injection, contrast media with low viscosities should be used. A further possibility to increase the IDR is warming-up the contrast medium to body temperature. (orig.) [de

Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

International Nuclear Information System (INIS)

Hsii, Y.H.

1975-01-01

The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The analytical model used for the program is described. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc. 11 references. (U.S.)

Optimum injection pressure of a cavitating jet on introduction of compressive residual stress into stainless steel

International Nuclear Information System (INIS)

Soyama, Hitoshi; Nagasaka, Kazuya; Takakuwa, Osamu; Naito, Akima

2011-01-01

In order to mitigate stress corrosion cracking of components used for nuclear power plants, introduction of compressive residual stress into sub-surface of the components is an effective maintenance method. The introduction of compressive residual stress using cavitation impact generated by injecting a high speed water jet into water was proposed. Water jet peening is now applying to reduce stress corrosion cracking of shrouds in the nuclear power plants. However, accidental troubles such as dropping off the components and cutting of the pipes by the jet occurred at the maintenance. In order to peen by the jet without damage, optimum injection pressure of the jet should be revealed. In the case of 'cavitation peening', cavitation is generated by injecting the high speed water jet into water. As working pressure at the cavitation peening is the pressure at cavitation bubble collapse, the injection pressure of the jet is not main parameter. The cavitation impact is increasing with the scale of the jet, i.e., scaling effect of the cavitation. It was revealed that the large scale jet at low injection pressure can introduce compressive residual stress into stainless steel comparing with the small scale jet at high injection pressure. As expected, a water jet at high injection pressure might make damage of the components. Namely, in order to avoid damage of the components, the jet at the low injection pressure will be suit for the introduction of compressive residual stress. In the present paper, in order to make clear optimum injection pressure of the cavitating jet for the introduction of compressive residual stress without damage, the residual stress of stainless steel treated by the jet at various injection pressure was measured by using an X-ray diffraction method. The injection pressure of the jet p 1 was varied from 5 MPa to 300 MPa. The diameter of the nozzle throat of the jet d was varied from 0.35 mm to 2.0 mm. The residual stress changing with depth was

Simulation of a loss of coolant accident with hydroaccumulator injection

International Nuclear Information System (INIS)

1988-10-01

An essential component of nuclear safety activities is the analysis of postulated accidents which are taken as a design basis for a facility. This analysis is usually carried out by using complex computer codes to simulate the behaviour of the plant and to calculate vital plant parameters, which are then compared with the design limits. Since these simulations cannot be verified at the plant itself, computer codes must be validated by comparing the results of calculations with experimental data obtained in test facilities. The IAEA, having identified the need for experimental data due to the difficulties of building integral test facilities and the high costs of these experiments, has accepted the offer of the Hungarian Academy of Sciences and organized two standard problem exercises. In these exercises, experimental data from the simulation of a 7.4% break loss of coolant accident was compared with analytical prediction of the behaviour of the facility calculated with computer codes. The second standard problem exercise involved a similar test, with the exception that in this case hydroaccumulator of the safety injection system were allowed to inject water in the system as anticipated in the design of the plant. This document presents a complete overview of the Second Standard Problem Exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many inter-related steps, therefore, no general conclusion on optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation. 22 refs, figs and tabs

LWR primary coolant pipe rupture test rig

International Nuclear Information System (INIS)

Yoshitoshi, Shyoji

1978-01-01

The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

A control-oriented approach to estimate the injected fuel mass on the basis of the measured in-cylinder pressure in multiple injection diesel engines

International Nuclear Information System (INIS)

Finesso, Roberto; Spessa, Ezio

2015-01-01

Highlights: • Control-oriented method to estimate injected quantities from in-cylinder pressure. • Able to calculate the injected quantities for multiple injection strategies. • Based on the inversion of a heat-release predictive model. • Low computational time demanding. - Abstract: A new control-oriented methodology has been developed to estimate the injected fuel quantities, in real-time, in multiple injection DI diesel engines on the basis of the measured in-cylinder pressure. The method is based on the inversion of a predictive combustion model that was previously developed by the authors, and that is capable of estimating the heat release rate and the in-cylinder pressure on the basis of the injection rate. The model equations have been rewritten in order to derive the injected mass as an output quantity, starting from use of the measured in-cylinder pressure as input. It has been verified that the proposed method is capable of estimating the injected mass of pilot pulses with an uncertainty of the order of ±0.15 mg/cyc, and the total injected mass with an uncertainty of the order of ±0.9 mg/cyc. The main sources of uncertainty are related to the estimation of the in-cylinder heat transfer and of the isentropic coefficient γ = c_p/c_v. The estimation of the actual injected quantities in the combustion chamber can represent a powerful means to diagnose the behavior of the injectors during engine operation, and offers the possibility of monitoring effects, such as injector ageing and injector coking, as well as of allowing an accurate control of the pilot injected quantities to be obtained; the latter are in fact usually characterized by a large dispersion, with negative consequences on the combustion quality and emission formation. The approach is characterized by a very low computational time, and is therefore suitable for control-oriented applications.

Direct vessel inclined injection system for reduction of emergency core coolant direct bypass in advanced reactors

International Nuclear Information System (INIS)

Yoon, Sang H.; Lee, Jong G.; Suh, Kune Y.

2006-01-01

Multidimensional thermal hydraulics in the APR1400 (Advanced Power Reactor 1400 MWe) downcomer during a large-break loss-of-coolant accident (LBLOCA) plays a pivotal role in determining the capability of the safety injection system. APR1400 adopts the direct vessel injection (DVI) method for more effective core penetration of the emergency core cooling (ECC) water than the cold leg injection (CLI) method in the OPR1000 (Optimized Power Reactor 1000 MWe). The DVI method turned out to be prone to occasionally lack in efficacious delivery of ECC to the reactor core during the reflood phase of a LBLOCA, however. This study intends to demonstrate a direct vessel inclined injection (DVII) method, one of various ideas with which to maximize the ECC core penetration and to minimize the direct bypass through the break during the reflood phase of a LBLOCA. The 1/7 scaled down THETA (Transient Hydrodynamics Engineering Test Apparatus) tests show that a vertical inclined nozzle angle of the DVII system increases the downward momentum of the injected ECC water by reducing the degree of impingement on the reactor downcomer, whereby lessening the extent of the direct bypass through the break. The proposed method may be combined with other innovative measures with which to ensure an enough thermal margin in the core during the course of a LBLOCA in APR1400

A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

International Nuclear Information System (INIS)

Siamidis, John; Mason, Lee

2006-01-01

This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H2O for the HRS pumped loop coolant working fluid. A detailed excel analytical model, HRS O pt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids

Requirements of coolants in nuclear reactors

International Nuclear Information System (INIS)

Abass, O. A. M.

2014-11-01

This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

The influence of thermal regime on gasoline direct injection engine performance and emissions

Science.gov (United States)

Leahu, C. I.; Tarulescu, S.

2016-08-01

This paper presents the experimental research regarding to the effects of a low thermal regime on fuel consumption and pollutant emissions from a gasoline direct injection (GDI) engine. During the experimental researches, the temperature of the coolant and oil used by the engine were modified 4 times (55, 65, 75 and 85 oC), monitoring the effects over the fuel consumption and emissions (CO2, CO and NOx). The variations in temperature of the coolant and oil have been achieved through AVL coolant and oil conditioning unit, integrated in the test bed. The obtained experimental results reveals the poor quality of exhaust gases and increases of fuel consumption for the gasoline direct injection engines that runs outside the optimal ranges for coolant and oil temperatures.

Peaking cladding temperature and break equivalent size of intermediate break loss of coolant accident

International Nuclear Information System (INIS)

Luo Bangqi

2012-01-01

The analysis results of intermediate break loss of coolant accident for the nuclear power plant of million kw level showed to be as following: (1) At the begin of life, the break occur simultaneity reactor shutdown with L(X)P. it's equivalent break size and peaking cladding temperature is respectively 20 cm and 849℃. (2) At the begin of life, the break occur simultaneity reactor shutdown without loop. the reactor coolant pumps will be stop after reactor shutdown 10 minutes, it's equivalent break size and peaking cladding temperature is respectively 10.5 cm and 921℃. (3) At the bur up of 31 GWd/t(EOC1). the break occur simultaneity reactor shutdown without loop, the reactor coolant pumps will be stop after reactor shutdown 20 minutes, it's equivalent break size and peaking cladding temperature is respectively 8 cm and 1145℃. The above analysis results showed that the peaking cladding temperature of intermediate break loss of coolant accident is not only related with the break equivalent size and core bur up, and is closely related with the stop time of coolant pumps because the coolant pumps would drive the coolant from safety system to produce the seal loop in break loop and affect the core coolant flow, results in the fuel cladding temperature increasing or damaging. Therefore, the break spectrum, burn up spectrum, the stop time of coolant pumps and operator action time will need to detail analysis and provide appropriate operating procedure, otherwise the peaking cladding temperature will exceed 1204℃ and threaten the safety of the reactor core when the intermediate break loss of coolant accident occur in some break equivalent size, burn up, stop pumps time and operator action not appropriate. The pressurizer pressure low signal simultaneity containment pressure higher signal were used as the operator manual close the signal of reactor coolant pumps after reactor shutdown of 20 minutes. have successful solved the operator intervention time from 10 minutes

Image processing analysis of combustion for D. I. diesel engine with high pressure fuel injection. ; Effects of air swirl and injection pressure. Nensho shashin no gazo shori ni yoru koatsu funsha diesel kikan no nensho kaiseki. ; Swirl oyobi funsha atsuryoku no eikyo

Energy Technology Data Exchange (ETDEWEB)

Yamaguchi, I. (Japan Automobile Research Institute, Inc., Tsukuba (Japan)); Tsujimura, K.

1994-02-25

This paper reports an image processing analysis of combustion for a high-pressure direct injection diesel engine on the effects of air swirl and injection pressure upon combustion in the diesel engine. The paper summarizes a method to derive gas flow and turbulence strengths, and turbulent flow mixing velocity. The method derives these parameters by detecting movement of brightness unevenness on two flame photographs through utilizing the mutual correlative coefficients of image concentrations. Five types of combustion systems having different injection pressures, injection devices, and swirl ratios were used for the experiment. The result may be summarized as follows: variation in the average value of the turbulent flow mixing velocities due to difference in the swirl ratio is small in the initial phase of diffusion combustion; the difference is smaller in the case of high swirl ratio than in the case of low swirl ratio after the latter stage of the injection; the average value is larger with the higher the injection pressure during the initial stage of the combustion; after termination of the injection, the value is larger in the low pressure injection; and these trends agree with the trend in the time-based change in heat generation rates measured simultaneously. 6 refs., 14 figs., 2 tabs.

Simulation of Spray Injection in the Pressurizer Using RELAP5

Directory of Open Access Journals (Sweden)

S. Dibyo

2017-08-01

Full Text Available A modeling research using Relap5 to assess the pressurizer of a pressurized water reactor(PWR power plant has been performed. The heater and water injection systems in the pressurizer system of the PWRare of greatimportance for system pressure control.The heater is designed to increase the pressure while the water sprayer injection is to perform depressurization. Most of studies conducted in the past mainly focused on determining the effects of nozzle spray design and droplet size using testing loops. The purpose of this simulation is to analyze the spray injection flow rate against the pressure characteristics of the pressurizer using RELAP5. Through this approach, the optimum injection flow rate of full scale plant pressurizer can be analyzed. The parameters investigated are pressure and temperature.In RELAP5, the pressurizer tank wasmodeled with six volume nodes and the heater was modeled by using heat structure. In the model, the sprayer takes water from the cold leg to inject it into the top of tank region.The resultsshowedthat the mass flow of about 4 kg/s is the mosteffectivevalueto limit pressure in the pressurizer to below 15.7 MPa. However, the flow rates of 8 kg/s and more cause overpressure. This simulation is usefulto complement the data related to the water flow rate injection systems of the pressurizer. Normal 0 false false false EN-US X-NONE X-NONE Organic coolants and their applications to fusion reactors

International Nuclear Information System (INIS)

Gierszewski, P.; Hollies, B.

1986-08-01

Organic coolants offer a unique set of characteristics for fusion applications. Their advantages include high-temperature (670 K or 400 degrees C) but low-pressure (2 MPa) operation, limited reactivity with lithium and lithium-lead, reduced corrosion and activation, good heat-transfer capabilities, no magnetohydrodynamic (MHD) effects, and an operating temperature range that extends to room temperature. The major disadvantages are decomposition and flammability. However, organic coolants have been extensively studied in Canada, including nineteen years with an operating 60-MW organic-cooled reactor. Proper attention to design and coolant chemistry controlled these potential problems to acceptable levels. This experience provides an extensive data base for design under fusion conditions. The organic fluid characteristics are described in sufficient detail to allow fusion system designers to evaluate organic coolants for specific applications. To illustrate and assess the potential applications, analyses are presented for organic-cooled blankets, first walls, high heat flux components and thermal power cycles. Designs are identified that take advantage of organic coolant features, yet have fluid decomposition related costs that are a small fraction of the overall cost of electricity. For example, organic-cooled first walls make lithium/ferritic steel blankets possible in high-field, high-surface-heat-flux tokamaks, and organic-cooled limiters (up to about 8 MW/m 2 surface heating) are a safer alternative to water cooling for liquid metal blanket concept. Organics can also be used in intermediate heat exchanger loops to provide efficient heat transfer with low reactivity and a large tritium barrier. 55 refs

Definition of the seventh dynamic AER benchmark-WWER-440 pressure vessel coolant mixing by re-connection of an isolated loop

International Nuclear Information System (INIS)

Kotsarev, A.; Lizorkin, M.; Petrin, R.

2010-01-01

The seventh dynamic benchmark is a continuation of the efforts to validate systematically codes for the estimation of the transient behavior of VVER type nuclear power plants. This benchmark is a continuation of the work in the sixth dynamic benchmark. It is proposed to be simulated the transient - re-connection of an isolated circulating loop with low temperature or low boron concentration in a VVER-440 plant. It is supposed to expand the benchmark to other cases when a different number of loops are in operation leading to different symmetric and asymmetric core boundary conditions. The purposes of the proposed benchmark are: 1) Best-estimate simulations of an transient with a coolant flow mixing in the Reactor Pressure Vessel of WWER-440 plant by re-connection of one coolant loop to the several ones on operation, 2) Performing of code-to-code comparisons. The core is at the end of its first cycle with a power of 1196.25 MWt. The basic additional difference of the 7-seventh benchmark is in the detailed description of the downcomer and bottom part of the reactor vessel that allow describing the effects of coolant mixing in the Reactor Pressure Vessel without any additional conservative assumptions. The burn-up and the power distributions at this reactor state have to be calculated by the participants. The thermohydraulic conditions of the core in the beginning of the transient are specified. Participants self-generated best estimate nuclear data is to be used. The main geometrical parameters of the plant and the characteristics of the control and safety systems are also specified. Use generated input data decks developed for a WWER-440 plant and for the applied codes should be used. The behaviour of the plant should be studied applying coupled system codes, which combine a three-dimensional neutron kinetics description of the core with a pseudo or real 3D thermohydraulics system code. (Authors)

Results of zinc injection test for Hamaoka Unit-1

International Nuclear Information System (INIS)

Kani, K.; Masuda, H.; Hayashi, Y.; Sudo, S.; Yamazaki, K.

1998-01-01

A zinc injection test was preformed at Hamaoka Nuclear Power Station Unit-1 for suppressing radiation dose rate on primary coolant recirculation piping after the replacement of piping. Zinc ion was injected by using injection system where Depleted Zinc Oxide was dissolved in carbonated water. Controllability of the system was sufficient to maintain concentration of zinc in primary water. The concentration of zinc in the primary coolant was controlled from 1 ppb to 5 ppb gradually. The increasing trend of concentration of Co-60 in the coolant was suppressed at zinc concentration of 3 ppb. It is evaluated that the deposition coefficient of Co-60 onto the surface of primary coolant recirculation piping was suppressed to one-third of previous cycle in average, and one-fourth of that just before injection start at zinc concentration of 5 ppb. We concluded that zinc injection is effective for suppressing dose rate on the primary coolant piping and no adverse effect occurs by zinc injection up to 5 ppb in the primary coolant. (J.P.N.)

Chaotic behavior of water column oscillator simulating pressure balanced injection system in passive safety reactor

International Nuclear Information System (INIS)

Morimoto, Y.; Madarame, H.; Okamoto, K.

2001-01-01

Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor called the System-integrated Pressurized Water Reactor (SPWR). In a loss of coolant accident, the Pressurizing Line (PL) and the Injection Line (IL) are passively opened. Vapor generated by residual heat pushes down the water level in the Reactor Vessel (RV). When the level is lower than the inlet of the PL, the vapor is ejected into the Containment Vessel (CV) through the PL. Then boronized water in the CV is injected into the RV through the IL by the static head. In an experiment using a simple apparatus, gas ejection and water injection were found to occur alternately under certain conditions. The gas ejection interval was observed to fluctuate considerably. Though stochastic noise affected the interval, the experimental results suggested that the large fluctuation was produced by an inherent character in the system. A set of piecewise linear differential equations was derived to describe the experimental result. The large fluctuation was reproduced in the analytical solution. Thus it was shown to occur even in a deterministic system without any source of stochastic noise. Though the derived equations simulated the experiment well, they had ten independent parameters governing the behavior of the solution. There appeared chaotic features and bifurcation, but the analytical model was too complicated to examine the features and mechanism of bifurcation. In this study, a new simple model is proposed which consists of a set of piecewise linear ordinary differential equations with only four independent parameters. (authors)

Numerical study of hot-leg ECC injection into the upper plenum of a pressurized water reactor

International Nuclear Information System (INIS)

Daly, B.J.; Torrey, M.D.; Rivard, W.C.

1981-01-01

In certain pressurized water reactor (PWR) designs, emergency core coolant (ECC) is injected through the hot legs into the upper plenum. The condensation of steam on this subcooled liquid stream reduces the pressure in the hot legs and upper plenum and thereby affects flow conditions throughout the reactor. In the present study, we examine countercurrent steam-water flow in the hot leg to determine the deceleration of the ECC flow that results from an adverse pressure gradient and from momentum exchange from the steam by interfacial drag and condensation. For the parameters examined in the study, water flow reversal is observed for a pressure drop of 22 to 32 mBar over the 1.5 m hot leg. We have also performed a three-dimensional study of subcooled water injection into air and steam environments of the upper plenum. The ECC water is deflected by an array of cylindrical guide tubes in its passage through the upper plenum. Comparisons of the air-water results with data obtained in a full scale experiment shows reasonable agreement, but indicates that there may be too much resistance to horizontal flow about the columns because of the use of a stair-step representation of the cylindrical guide tube cross section. Calculations of flow past single columns of stair-step, square and circular cross section do indicate excessive water deeentrainment by the noncircular column. This has prompted the use of an arbitrary mesh computational procedure to more accuratey represent the circular cross-section guide tubes. 15 figures

Modelling the effect of injection pressure on heat release parameters and nitrogen oxides in direct injection diesel engines

Directory of Open Access Journals (Sweden)

Yüksek Levent

2014-01-01

Full Text Available Investigation and modelling the effect of injection pressure on heat release parameters and engine-out nitrogen oxides are the main aim of this study. A zero-dimensional and multi-zone cylinder model was developed for estimation of the effect of injection pressure rise on performance parameters of diesel engine. Double-Wiebe rate of heat release global model was used to describe fuel combustion. extended Zeldovich mechanism and partial equilibrium approach were used for modelling the formation of nitrogen oxides. Single cylinder, high pressure direct injection, electronically controlled, research engine bench was used for model calibration. 1000 and 1200 bars of fuel injection pressure were investigated while injection advance, injected fuel quantity and engine speed kept constant. The ignition delay of injected fuel reduced 0.4 crank angle with 1200 bars of injection pressure and similar effect observed in premixed combustion phase duration which reduced 0.2 crank angle. Rate of heat release of premixed combustion phase increased 1.75 % with 1200 bar injection pressure. Multi-zone cylinder model showed good agreement with experimental in-cylinder pressure data. Also it was seen that the NOx formation model greatly predicted the engine-out NOx emissions for both of the operation modes.

Injection Pressure as a Marker of Intraneural Injection in Procedures of Peripheral Nerves Blockade

Directory of Open Access Journals (Sweden)

Ilvana Vučković

2006-11-01

Full Text Available The blockade of peripheral nerves carries a certain risk of unwanted complications, which can follow after unintentional intraneural injection of a local anesthetic. Up till today, the research of measuring injection pressure has been based on animal models, even though the histological structure of periphery nerve is different for animal and human population, so the application pressure which presages intraneural injection in human population is still unknown. As material in performing this study there have been used 12 Wistar rats and 12 delivered stillborns. After bilateral access to the median nerve, we applied 3 ml of 2% lidocaine with epinephrine, with the help of automatic syringe charger. The needle was at first placed perineural on one side, and then intraneural on the other side of both examination groups. During every application the pressure values were monitored using the manometer, and then they were analyzed by special software program BioBench. All perineural injections resulted with the pressure < or = 27.92 kPa, while the majority of intraneural injections were combined with the injectionpressure > or = 69.8 kPa. The difference between intraneural and perineural injection pressures for the two different examination groups (rats and delivered stillborns was not statistically significant (P>0.05. As prevention from intraneural injections today are in use two methods: the method of causing paresthesia or the method of using the peripheral nerve stimulator. However the nerve injury can still occur, independent from the technique used. If our results are used in clinical practice on human population, than the high injection pressure could be the markerof intraneural lodging of a needle.

Modulated diesel fuel injection strategy for efficient-clean utilization of low-grade biogas

International Nuclear Information System (INIS)

Wang, Xiaole; Qian, Yong; Zhou, Qiyan; Lu, Xingcai

2016-01-01

Highlights: • Influences of direct injection strategy on biogas RCCI mode are researched. • Excessive early pilot injection timing leads to the retard of combustion. • Overall indicated thermal efficiency of low-grade biogas can be higher than 40%. • Pilot injection timing has strong influence on particle size distribution. • Composition of biogas has a great influence on the gas emissions. - Abstract: Recently, as a kind of renewable fuel, low-grade biogas has been researched to apply in internal combustion engine. In this paper, an experimental study was conducted to study the influence of injection strategies on the efficient utilization of low-grade biogas in Reactivity Controlled Compression Ignition (RCCI) mode with port fuel injection of biogas and in-cylinder direct injection of diesel based on a modified electronic controlled high-pressure directly injected compression ignition engine. Considered the high proportion of inert gas in biogas, a four-components simulated gas (H_2:CO:CH_4:N_2 = 5:40:5:50 vol%) has been selected as test fuels to simulate biogas. The effects of several injection control parameters such as pilot injection timing, main injection timing, common rail pressure and pilot injection ratio on the combustion and emissions are analyzed in detail. The research demonstrates that the main injection timing can effectively control the combustion phase and excessive early pilot injection timing leads to retard of combustion. CO emissions are relatively high due to homogenous charge of biogas. NOx and smoke emissions can be effectively controlled. In RCCI mode, the indicated thermal efficiency of biogas/diesel can reach 40%.

Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

International Nuclear Information System (INIS)

Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

1995-01-01

The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

Management of large scale coolant channel replacement programme for Indian PHWRs

International Nuclear Information System (INIS)

Bhatnagar, V.K.; Chadda, S.K.; Arya, R.C.

1994-01-01

Coolant channel assemblies form most important core components of pressurised heavy water reactors. Zirconium alloy pressure tube which form part of coolant channel assemblies are subjected to environment of high neutron flux, high pressure and temperature. Under those operating environmental conditions, the pressure tubes material undergoes degradation of metallurgical and mechanical properties in addition to dimensional changes. The coolant channels are subjected to an in-service inspection (ISI) programme for monitoring the health particularly of the pressure tubes. The en-mass replacement of pressure tubes is needed after most of the pressure tubes show unacceptable conditions for an assured safe and reliable operation. An overview of various issues pertaining to this aspect is presented. (author). 4 figs

Stress corrosion cracking studies on ferritic low alloy pressure vessel steel - water chemistry and modelling aspects

International Nuclear Information System (INIS)

Tipping, P.; Ineichen, U.; Cripps, R.

1994-01-01

The susceptibility of low alloy ferritic pressure vessel steels (A533-B type) to stress corrosion cracking (SCC) degradation has been examined using various BWR type coolant chemistries. Fatigue pre-cracked wedge-loaded double cantilever beams and also constantly loaded 25 mm thick compact tension specimens have shown classical SCC attack. The influence of parameters such as dissolved oxygen content, water impurity level and conductivity, material chemical composition (sulphur content) and stress intensity level are discussed. The relevance of SCC as a life-limiting degradation mechanism for low alloy ferritic nuclear power plant PV steel is examined. Some parameters, thought to be relevant for modelling SCC processes in low alloy steels in simulated BWR-type coolant, are discussed. 8 refs., 1 fig., 4 tabs

Analyses of Decrease in Reactor Coolant Flow Rate in SMART

International Nuclear Information System (INIS)

Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

2011-01-01

SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

Analysis of events related to cracks and leaks in the reactor coolant pressure boundary

Energy Technology Data Exchange (ETDEWEB)

Ballesteros, Antonio, E-mail: Antonio.Ballesteros-Avila@ec.europa.eu [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Sanda, Radian; Peinador, Miguel; Zerger, Benoit [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Negri, Patrice [IRSN: Institut de Radioprotection et de Sûreté Nucléaire (France); Wenke, Rainer [GRS: Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH (Germany)

2014-08-15

Highlights: • The important role of Operating Experience Feedback is emphasised. • Events relating to cracks and leaks in the reactor coolant pressure boundary are analysed. • A methodology for event investigation is described. • Some illustrative results of the analysis of events for specific components are presented. - Abstract: The presence of cracks and leaks in the reactor coolant pressure boundary may jeopardise the safe operation of nuclear power plants. Analysis of cracks and leaks related events is an important task for the prevention of their recurrence, which should be performed in the context of activities on Operating Experience Feedback. In response to this concern, the EU Clearinghouse operated by the JRC-IET supports and develops technical and scientific work to disseminate the lessons learned from past operating experience. In particular, concerning cracks and leaks, the studies carried out in collaboration with IRSN and GRS have allowed to identify the most sensitive areas to degradation in the plant primary system and to elaborate recommendations for upgrading the maintenance, ageing management and inspection programmes. An overview of the methodology used in the analysis of cracks and leaks related events is presented in this paper, together with the relevant results obtained in the study.

Pressure thermal shock analysis for nuclear reactor pressure vessel

International Nuclear Information System (INIS)

Galik, G.; Kutis, V.; Jakubec, J.; Paulech, J.; Murin, J.

2015-01-01

The appearance of structural weaknesses within the reactor pressure vessel or its structural failure caused by crack formation during pressure thermal shock processes pose as a severe environmental hazard. Coolant mixing during ECC cold water injection was simulated in a detailed CFD analysis. The temperature distribution acting on the pipe wall internal surface was calculated. Although, the results show the formation of high temperature differences and intense gradients, an additional structural analysis is required to determine the possibility of structural damage from PTS. Such an analysis will be the subject of follow-up research. (authors)

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

CANDU with supercritical water coolant: conceptual design features

International Nuclear Information System (INIS)

Spinks, N.

1997-01-01

An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

Device for preventing coolant in a reactor from being lost

International Nuclear Information System (INIS)

Maruyama, Hiromi; Matsumoto, Tomoyuki.

1975-01-01

Object: To prevent all of coolant from being lost from the core at the time of failure in rupture of pipe in a recirculation system to cool the core with the coolant remained within the reactor. Structure: A valve, which will be closed when a water level of the coolant within the core is in a level less than a predetermined level, is provided on a recirculating water outlet nozzle in a pressure vessel to thereby prevent the coolant from being lost when the pipe is broken, thus cooling the core by means of reduced-pressure boiling of coolant remained within the core and boiling due to heat, and restraining core reactivity by means of void produced at that time. (Kamimura, M.)

Fuel coolant interaction experiment by direct electrical heating method

International Nuclear Information System (INIS)

Takeda, Tsuneo; Hirano, Kenmei

1979-01-01

In the PCM (Power Cooling Mismatch) experiments, the FCI (Fuel Coolant Interaction) test is one of necessary tests in order to predict various phenomena that occur during PCM in the core. A direct electrical heating method is used for the FCI tests for fuel pellet temperature of over 1000 0 C. Therefore, preheating is required before initiating the direct electrical heating. The fuel pin used in the FCI tests is typical LWR fuel element, which is surrounded by coolant water. It is undersirable to heat up the coolant water during preheating of the fuel pin. Therefore, a zirconia (ZrO 2 ) pellet which is similar to a UO 2 pellet in physical and chemical properties is used. Electric property (electric conductivity) of ZrO 2 is particularly suitable for direct electrical heating as in the case of UO 2 . In this experiment, ZrO 2 pellet (melting point 2500 0 C) melting was achieved by use of both preheating and direct electrical heating. Temperature changes of coolant and fuel surface, as well as the pressure change of coolant water, were measured. The molten fuel interacted with the coolant and generated shock waves. A portion of this molten fuel fragmented into small particles during this interaction. The peak pressure of the observed shock wave was about 35 bars. The damaged fuel pin was photographed after disassembly. This report shows the measured coolant pressure changes and the coolant temperature changes, as well as photographs of damaged fuel pin and fuel fragments. (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

Diesel Combustion and Emission Using High Boost and High Injection Pressure in a Single Cylinder Engine

Science.gov (United States)

Aoyagi, Yuzo; Kunishima, Eiji; Asaumi, Yasuo; Aihara, Yoshiaki; Odaka, Matsuo; Goto, Yuichi

Heavy-duty diesel engines have adopted numerous technologies for clean emissions and low fuel consumption. Some are direct fuel injection combined with high injection pressure and adequate in-cylinder air motion, turbo-intercooler systems, and strong steel pistons. Using these technologies, diesel engines have achieved an extremely low CO2 emission as a prime mover. However, heavy-duty diesel engines with even lower NOx and PM emission levels are anticipated. This study achieved high-boost and lean diesel combustion using a single cylinder engine that provides good engine performance and clean exhaust emission. The experiment was done under conditions of intake air quantity up to five times that of a naturally aspirated (NA) engine and 200MPa injection pressure. The adopted pressure booster is an external supercharger that can control intake air temperature. In this engine, the maximum cylinder pressure was increased and new technologies were adopted, including a monotherm piston for endurance of Pmax =30MPa. Moreover, every engine part is newly designed. As the boost pressure increases, the rate of heat release resembles the injection rate and becomes sharper. The combustion and brake thermal efficiency are improved. This high boost and lean diesel combustion creates little smoke; ISCO and ISTHC without the ISNOx increase. It also yields good thermal efficiency.

Post injection pressures in well treatments

Energy Technology Data Exchange (ETDEWEB)

Shaw, G

1967-06-05

Behavior of wellhead pressure immediately after injection of liquids or slurries in well completion and workover treatments can often indicate the success of the operation. Since the rate of wellhead pressure build-down after injection is related to the permeability of the exposed formation to the treating fluid, interpretation of success or failure of the fluid to communicate with the reservoir is possible. Treatments designed to plug-up or clean-out formation flow channels can both be evaluated. Early appreciation can speed completion and workover operations. An explanation of the phenomena of increasing bottomhole treating pressure during fracture-type treatments, and the change in it throughout the life of a well, will result in better understanding of basic fracturing mechanics. On-the-job observations of decreasing rate of pressure build-down after increments of stage squeeze cementing will help the well-site engineer to vary the volume of increments of slurry and the duration of each stage.

Coolant leakage detecting device

International Nuclear Information System (INIS)

Yamauchi, Kiyoshi; Kawai, Katsunori; Ishihara, Yoshinao.

1995-01-01

The device of the present invention judges an amount of leakage of primary coolants of a PWR power plant at high speed. Namely, a mass of coolants contained in a pressurizer, a volume controlling tank and loop regions is obtained based on a preset relational formula and signals of each of process amount, summed up to determine the total mass of coolants for every period of time. The amount of leakage for every period of time is calculated by a formula of Karman's filter based on the total mass of the primary coolants for every predetermined period of time, and displays it on CRT. The Karman's filter is formed on every formula for several kinds of states formed based on the preset amount of the leakage, to calculate forecasting values for every mass of coolants. An adaptable probability for every preset leakage amount is determined based on the difference between the forecast value and the observed value and the scattering thereof. The adaptable probability is compared with a predetermined threshold value, which is displayed on the CRT. This device enables earlier detection of leakage and identification of minute leakage amount as compared with the prior device. (I.S.)

Pressure tube reactor

International Nuclear Information System (INIS)

Susuki, Akira; Murata, Shigeto; Minato, Akihiko.

1993-01-01

In a pressure tube reactor, a reactor core is constituted by arranging more than two units of a minimum unit combination of a moderator sealing pipe containing a calandria tube having moderators there between and a calandria tube and moderators. The upper header and a lower header of the calandria tank containing moderators are communicated by way of the moderator sealing tube. Further, a gravitationally dropping mechanism is disposed for injecting neutron absorbing liquid to a calandria gas injection portion. A ratio between a moderator volume and a fuel volume is defined as a function of the inner diameter of the moderator sealing tube, the outer diameter of the calandria tube and the diameter of fuel pellets, and has no influence to intervals of a pressure tube lattice. The interval of the pressure tube lattice is enlarged without increasing the size of the pressure tube, to improve production efficiency of the reactor and set a coolant void coefficient more negative, thereby enabling to improve self controllability and safety. Further, the reactor scram can be conducted by injecting neutron absorbing liquid. (N.H.)

Improvement of lifetime availability through design, inspection, repair and replacement of coolant channels of Indian Pressurized Heavy Water Reactors

International Nuclear Information System (INIS)

Rupani, B.B.; Sinha, R.K.

1998-01-01

This paper covers an overview of the work carried out for the life management of the coolant channels of Indian Pressurised Heavy Water Reactors. In order to improve maintainability of the coolant channels and reduce down time needed for periodical creep adjustment, improved design of channel hardware were developed. The modular insulation panel, designed as a substitute for the jig saw panels, reduces the time needed for accessing the space around the end-fitting significantly. A compact mechanical snubber has been developed to totally eliminate the need for periodic creep adjustment. In addition, the paper also describes the technologies developed for performing some special inspection, repair and replacement tasks for the coolant channels. These include systems for garter spring repositioning by Mechanical Flexing Technique for fresh reactors and Integrated Garter Spring Repositioning System for operating reactors. A tooling system, developed for in-situ retrieval of sliver scrape samples from pressure tubes, is also described. These samples can be analysed in laboratories to yield valuable information on hydrogen concentration in pressure tube material. The current and planned activities towards development of technologies for improvement of the life time availability of the power plants are addressed. (author)

Computational Fluid Dynamics Analysis of High Injection Pressure Blended Biodiesel

Science.gov (United States)

Khalid, Amir; Jaat, Norrizam; Faisal Hushim, Mohd; Manshoor, Bukhari; Zaman, Izzuddin; Sapit, Azwan; Razali, Azahari

2017-08-01

Biodiesel have great potential for substitution with petrol fuel for the purpose of achieving clean energy production and emission reduction. Among the methods that can control the combustion properties, controlling of the fuel injection conditions is one of the successful methods. The purpose of this study is to investigate the effect of high injection pressure of biodiesel blends on spray characteristics using Computational Fluid Dynamics (CFD). Injection pressure was observed at 220 MPa, 250 MPa and 280 MPa. The ambient temperature was kept held at 1050 K and ambient pressure 8 MPa in order to simulate the effect of boost pressure or turbo charger during combustion process. Computational Fluid Dynamics were used to investigate the spray characteristics of biodiesel blends such as spray penetration length, spray angle and mixture formation of fuel-air mixing. The results shows that increases of injection pressure, wider spray angle is produced by biodiesel blends and diesel fuel. The injection pressure strongly affects the mixture formation, characteristics of fuel spray, longer spray penetration length thus promotes the fuel and air mixing.

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

Influence of coolant motion on structure of hardened steel element

Directory of Open Access Journals (Sweden)

A. Kulawik

2008-08-01

Full Text Available Presented paper is focused on volumetric hardening process using liquid low melting point metal as a coolant. Effect of convective motion of the coolant on material structure after hardening is investigated. Comparison with results obtained for model neglecting motion of liquid is executed. Mathematical and numerical model based on Finite Element Metod is described. Characteristic Based Split (CBS method is used to uncouple velocities and pressure and finally to solve Navier-Stokes equation. Petrov-Galerkin formulation is employed to stabilize convective term in heat transport equation. Phase transformations model is created on the basis of Johnson-Mehl and Avrami laws. Continuous cooling diagram (CTPc for C45 steel is exploited in presented model of phase transformations. Temporary temperatures, phases participation, thermal and structural strains in hardening element and coolant velocities are shown and discussed.

Effect of Intra Vitreal Injection of Bevacizumab on Intra-Occular Pressure

International Nuclear Information System (INIS)

Jaffar, S.; Tayyab, A.; Matin, Z. I.; Masrur, A.; Naqaish, R.

2016-01-01

Background: Bevacizumab has been in use as a therapeutic agent for macular oedema for several years. While its efficacy has been well documented, its use has been shown to cause a transient rise in the intra-ocular pressure. The aim of this study was to evaluate the long term effect of intra-vitreal injection of Bevacizumab on Intra-ocular pressure. Methods: One hundred eyes (n=100) of one hundred patients, requiring intra-vitreal injection of Bevacizumab for diabetic macular oedema were recruited from Shifa Foundation Community Health Centre (SFCHC) between January and December 2014. Patients of glaucoma, ocular hyper-tension, known allergy to Bevacizumab or had injections of Bevacizumab prior to the study were excluded. Intra-ocular pressure was measured using a Goldmann applanation tonometer, prior to, and at six and twelve months after the injection. The pre- and post- injection Intra-ocular pressure was entered into the database. Test of significance was applied to investigate whether there was a significant change in intra-ocular pressure after the injection. Results: The mean age of the patient was 56.97 years (±14.97). The mean intra-ocular pressure was 13.86 (±3.16) mmHg before injection, while post-injection mean Intra-Ocular pressure was 14.21 (±3.12) mmHg and 13.79 (±3.07) at six and twelve months respectively. Between baseline and six months there was a statistically significant difference in intra-ocular pressure (p=0.03), while no significant difference existed in the intra-ocular pressure between baseline and twelve months (p=0.92). Conclusion: Intra-vitreal injection of Bevacizumab is associated with a statically significant rise in intra-ocular pressure at six months, while no significant difference was seen at twelve months compared to baseline. (author)

Construction of a Direct Water-Injected Two-Stroke Engine for Phased Direct Fuel Injection-High Pressure Charging Investigations

Science.gov (United States)

Somsel, James P.

1998-01-01

The development of a water injected Orbital Combustion Process (OCP) engine was conducted to assess the viability of using the powerplant for high altitude NASA aircraft and General Aviation (GA) applications. An OCP direct fuel injected, 1.2 liter, three cylinder, two-stroke engine has been enhanced to independently inject water directly into the combustion chamber. The engine currently demonstrates low brake specific fuel consumption capability and an excellent power to weight ratio. With direct water injection, significant improvements can be made to engine power, to knock limits/ignition advance timing, and to engine NO(x) emissions. The principal aim of the testing was to validate a cyclic model developed by the Systems Analysis Branch at NASA Ames Research Center. The work is a continuation of Ames' investigations into a Phased Direct Fuel Injection Engine with High Pressure Charging (PDFI-ITPC).

Continuous surveillance of reactor coolant circuit integrity

International Nuclear Information System (INIS)

1986-01-01

Continuous surveillance is important to assuring the integrity of a reactor coolant circuit. It can give pre-warning of structural degradation and indicate where off-line inspection should be focussed. These proceedings describe the state of development of several techniques which may be used. These involve measuring structural vibration, core neutron noise, acoustic emission from cracks, coolant leakage, or operating parameters such as coolant temperature and pressure. Twenty three papers have been abstracted and indexed separately for inclusion in the data base

High pressure injection of dimethyl ether

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

The purpose of this investigation was to achieve a better understanding of the fundamental spray behavior of DME (Dimenthyl Ether) using a standard diesel pump with pintle and hole nozzles. Fundamental spray behavior was characterized by determining fuel spray penetration and angle, atomization and evaporation. The influences of opening pressure, nozzle geometry and ambient pressure above and below the critical pressure of the fuel on the spray behavior were investigated. The influence of opening pressures on the spray characteristics for the hole nozzle was investigated. The results showed that for opening pressures of 120 bar and 180 bar the spray has a similar appearance. For the higher opening pressure (200 bar and 240 bar), the initial spray breaks up very rapidly giving a high initial spray angle. The opening pressure had little influence on spray penetration. The spray angle later in the injection increased as the opening pressure was decreased. Above the critical pressure, the spray from the hole nozzle had a more irregular shape. Penetration decreased and the spray angle increased above the critical pressure. Three pintle nozzles with different geometries and opening pressures were tested. The appearance of the three sprays were very similar. The sprays seemed to be more sharply pointed as the nozzle hole angle decreased. The nozzle with the 4 deg. hole nozzle angle and an opening pressure of 280 bar had the highest penetration and highest initial spray angle. The pintle nozzle with the 12 deg. hole nozzle angle and opening pressure of approx. 450 bar was tested above the critical ambient pressure. Penetration was very similar for injection above and below the critical ambient pressure, while the spray angle decreased for the spray above the critical ambient pressure. (au)

Analysis of a water-coolant leak into a very high-temperature vitrification chamber

International Nuclear Information System (INIS)

Felicione, F. S.

1998-01-01

A coolant-leakage incident occurred during non-radioactive operation of the Plasma Hearth Process waste-vitrification development system at Argonne National Laboratory when a stray electric arc ruptured az water-cooling jacket. Rapid evaporation of the coolant that entered the very high-temperature chamber pressurized the normally sub-atmospheric system above ambient pressure for over 13 minutes. Any positive pressurization, and particularly a lengthy one, is a safety concern since this can cause leakage of contaminants from the system. A model of the thermal phenomena that describe coolant/hot-material interactions was developed to better understand the characteristics of this type of incident. The model is described and results for a variety of hypothetical coolant-leak incidents are presented. It is shown that coolant leak rates above a certain threshold will cause coolant to accumulate in the chamber, and evaporation from this pool can maintain positive pressure in the system long after the leak has been stopped. Application of the model resulted in reasonably good agreement with the duration of the pressure measured during the incident. A closed-form analytic solution is shown to be applicable to the initial leak period in which the peak pressures are generated, and is presented and discussed

One-phase and two-phase homologous curves for coolant pumps of the pressurized light water nuclear reactors

International Nuclear Information System (INIS)

Santos, G.A. dos.

1990-01-01

The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The single-phase pump characteristics are an essential feature for operational transients studies, for example, the shut-down and start-up of pump. These parameters, in terms of the homologous curves, set up the complete performance of the pump and are input for transients and accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the single-phase and two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

Failure rates in Barsebaeck-1 reactor coolant pressure boundary piping. An application of a piping failure database

International Nuclear Information System (INIS)

Lydell, B.

1999-05-01

This report documents an application of a piping failure database to estimate the frequency of leak and rupture in reactor coolant pressure boundary piping. The study used Barsebaeck-1 as reference plant. The study tried two different approaches to piping failure rate estimation: 1) PSA-style, simple estimation using Bayesian statistics, and 2) fitting of statistical distribution to failure data. A large, validated database on piping failures (like the SKI-PIPE database) supports both approaches. In addition to documenting leak and rupture frequencies, the SKI report describes the use of piping failure data to estimate frequency of medium and large loss of coolant accidents (LOCAs). This application study was co sponsored by Barsebaeck Kraft AB and SKI Research

Hydrodynamics of Safety Injection Tank with Fluidic Device in Recent Regulation

International Nuclear Information System (INIS)

Bang, Young Seok; Yoo, Seung Hun

2016-01-01

Safety Injection Tank (SIT) with Fluidic Device (FD) has been used in several APR1400 nuclear power plants. It was designed to provide a longer passive safety injection than the existing accumulator to improve the safety for Large Break Loss-of-Coolant Accident (LBLOCA) by changing the injected flow through the FD and the standpipe of the SIT. As a result, high flow injection phase and the subsequent low flow one can be achieved as longer than the existing accumulator. The present paper discusses the major concerns related to SIT hydrodynamics and the directions to resolution recently concerned. Modeling of SIT/FD by total hydraulic resistances, potential of nitrogen intrusion, and effect of initial pressure of SIT testing are included. Based on the discussion, a table of the important phenomena of the SIT/FD was proposed with the relevancy of the calculation models applied. The present paper discussed the SIT hydrodynamics including the modeling of SIT/FD by total hydraulic resistances, potential of nitrogen intrusion, and effect of initial pressure of SIT testing. Also a table of the important phenomena of the SIT/FD was proposed with the relevancy of the calculation models applied. The following conclusions are obtained uncertainty due to the assumption of the total Kfactor as constant for high flow, transition phase, and low flow phase should be considered and nitrogen intrusion phenomena during the transition phase should be considered with a conservatism, especially considering the current situation of nonmeasuring the standpipe level

Hydrodynamics of Safety Injection Tank with Fluidic Device in Recent Regulation

Energy Technology Data Exchange (ETDEWEB)

Bang, Young Seok; Yoo, Seung Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

Safety Injection Tank (SIT) with Fluidic Device (FD) has been used in several APR1400 nuclear power plants. It was designed to provide a longer passive safety injection than the existing accumulator to improve the safety for Large Break Loss-of-Coolant Accident (LBLOCA) by changing the injected flow through the FD and the standpipe of the SIT. As a result, high flow injection phase and the subsequent low flow one can be achieved as longer than the existing accumulator. The present paper discusses the major concerns related to SIT hydrodynamics and the directions to resolution recently concerned. Modeling of SIT/FD by total hydraulic resistances, potential of nitrogen intrusion, and effect of initial pressure of SIT testing are included. Based on the discussion, a table of the important phenomena of the SIT/FD was proposed with the relevancy of the calculation models applied. The present paper discussed the SIT hydrodynamics including the modeling of SIT/FD by total hydraulic resistances, potential of nitrogen intrusion, and effect of initial pressure of SIT testing. Also a table of the important phenomena of the SIT/FD was proposed with the relevancy of the calculation models applied. The following conclusions are obtained uncertainty due to the assumption of the total Kfactor as constant for high flow, transition phase, and low flow phase should be considered and nitrogen intrusion phenomena during the transition phase should be considered with a conservatism, especially considering the current situation of nonmeasuring the standpipe level.

CAREM-25: considerations about primary coolant chemistry

International Nuclear Information System (INIS)

Chocron, Mauricio; Iglesias, Alberto M.; Raffo Calderon, Maria C.; Villegas, Marina

2000-01-01

World operating experience, in conjunction with basic studies has been modifying chemistry specifications for the primary coolant of water cooled nuclear reactors along with the reactor type and structural materials involved in the design. For the reactor CAREM-25, the following sources of information have been used: 1) Experience gained by the Chemistry Department of the National Atomic Energy Commission (CNEA, Argentina); 2) Participation of the Chemistry Department (CNEA) in international cooperation projects; 3) Guidelines given by EPRI, Siemens-KWU, AECL, etc. Given the main objectives: materials integrity, low radiation levels and personnel safety, which are in turn a balance between the lowest corrosion and activity transport achievable and considering that the CAREM-25 is a pressurized vessel integrated reactor, a group of guidelines for the chemistry and additives for the primary coolant have been given in the present work. (author)

Low-activation lead coolant for advanced small modular NPP

International Nuclear Information System (INIS)

Khorasanov, G.L.; Ivanov, A.P.; Blokhin, A.I.

2001-01-01

The purpose of the paper is in studying perspectives of a new heavy liquid metal coolant for a small fast reactor (FR) concept. To reduce the post irradiation activity of the coolant the using of lead isotope, Pb-206, instead of natural lead, Pb-nat, is offered. In this case the accumulation of such hazardous radionuclides, as Po-210, Bi-208, Bi-207, essentially decreases. The interval of the lead-206 coolant cost which does not exceed 20% of the overall FR cost is estimated. The possibility of lead-206 obtaining for FR needs with the centrifugal separation technique is pointed out. (author)

Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

International Nuclear Information System (INIS)

Hsii, Y.H.

1976-06-01

The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The report describes the analytical model used for the program. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The user is required to input the description of the discharge of coolant, the boiling of residual water by reactor decay heat, the superheating of steam passing through the core, and metal-water reactions. The reactor building is separated into liquid and vapor regions. Each region is in thermal equilibrium itself, but the two may not be in thermal equilibrium; the liquid and gaseous regions may have different temperatures. The reactor building is represented as consisting of several heat-conducting structures whose thermal behavior can be described by the one-dimensional multi-region heat conduction equation. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc

Atucha I nuclear power plant: Probabilistic safety study. Loss-of-coolant accidents

International Nuclear Information System (INIS)

Perez, S.S.

1987-01-01

The plant response to the group of events 'large coolant loss' in order to evaluate the associated risk is analyzed. The event that covers all events of similar sequence due to its evolution features, being also the most demanded, is selected as starting event. The representative event is the 'guillotine type rupture of cold primary branch'. An annual occurrence frequency of 10/year is assumed for this event. The safety systems, when the event occurs, must assure the reactor shutdown and the core cooling, creating a heat sink to remove the decay heat. The annual frequency of core meltdown due to great loss of coolant is obtained multiplying the annual frequency of the starting event by the probability of failure of involved safety systems. By means of failure trees, the following is obtained: a) probability of failure to demand of the boron injection shutdown system = 4 x 10 -2 ; b) probability of failure to demand of the high pressure safety injection = 3 x 10 -3 ; c) probability of emergency cooling system failure = 4.4 x 10 -2 . Therefore, the three possible sequences of core meltdown have the following frequencies: λ 1 = 4 x 10 -6 /year λ 2 = 3 x 10 -7 /year λ 3 = 4.4 x 10 -6 /year. (Author)

Reactor coolant cleanup device

International Nuclear Information System (INIS)

Igarashi, Noboru.

1986-01-01

Purpose: To enable to introduce reactor water at high temperature and high pressure as it is, as well as effectively adsorb to eliminate cobalt in reactor water. Constitution: The coolant cleanup device comprises a vessel main body inserted to coolant pipeway circuits in a water cooled reactor power plant and filters contained within the vessel main body. The filters are prepared by coating and baking powder of metal oxides such as manganese ferrite having a function capable of adsorbing cobalt in the coolants onto the surface of supports made of metals or ceramics resistant to strong acids and alkalies in the form of three-dimensional network structure, for example, zircaloy-2, SUS 303 and the zirconia (baking) to form a basic filter elements. The basic filter elements are charged in plurality to the vessel main body. (Kawaiami, Y.)

Loss of Coolant Accidents (LOCA): Study of CAREM Reactor Response

International Nuclear Information System (INIS)

Gonzalez, Jose; Gimenez, Marcelo

2000-01-01

We analyzed the neutronic and thermohydraulic response of CAREM25 reactor and the safety systems involved in a Loss Of Coolant Accident (LOCA).This parametric analysis considers several break diameters (1/2inch, 3/4inch, 1inch, 1.1/2inch and 2inches) in the vapor zone of the Reactor Pressure Vessel.For each accidental sequence, the successful operation of the following safety systems is modeled: Second Safety System (SSS), Residual Heat Removal System (RHRS) and Safety Injection System (SIS). Availability of only one module is postulated for each system.On the other hand, the unsuccessful operation of all safety systems is postulated for each accidental sequence.In both cases the First Shutdown System (FSS) actuates, and the loss of Steam Generator secondary flow and Chemical and Control of Volume System (CCVS) unavailability are postulated.Maximum loss of coolant flow, reactor power and time for safety systems operation are analyzed, as well as its set point parameters.We verified that safety systems are dimensioned to satisfy the 48 hours cooling criteria

Containment for low temperature district nuclear-heating reactor

International Nuclear Information System (INIS)

He Shuyan; Dong Duo

1992-03-01

Integral arrangement is adopted for Low Temperature District Nuclear-heating Reactor. Primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with reactor core. Primary coolant flows through reactor core and primary heat exchangers in natural circulation. Primary coolant pipes penetrating the wall of reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of pressure boundary of primary coolant. Therefore the small sized metallic containment closed to the wall of reactor vessel can be used for the reactor. Design principles and functions of the containment are as same as the containment for PWR. But the adoption of small sized containment brings about some benefits such as short period of manufacturing, relatively low cost, and easy for sealing. Loss of primary coolant accident would not be happened during the rupture accident of primary coolant pressure boundary inside the containment owing to its intrinsic safety

High converter pressurized water reactor with heavy water as a coolant

International Nuclear Information System (INIS)

Ronen, Y.; Reyev, D.

1983-01-01

There is an increasing interest in water breeder and high converter reactors. The increase in the conversion ratio of these reactors is obtained by hardening the neutron spectrum achieved by tightening the reactor's lattice. Another way of hardening the neutron spectrum is to replace the light water with heavy water. Two pressurized water reactor fuel cycles that use heavy water as a coolant are considered. The first fuel cycle is based on plutonium and depleted uranium, and the second cycle is based on plutonium and enriched uranium. The uranium ore and separative work unit (SWU) requirements are calculated as well as the fuel cycle cost. The savings in uranium ore are about40 and 60% and about40% in SWU for both fuel cycles considered

Pressure drops in low pressure local boiling

International Nuclear Information System (INIS)

Courtaud, Michel; Schleisiek, Karl

1969-01-01

For prediction of flow reduction in nuclear research reactors, it was necessary to establish a correlation giving the pressure drop in subcooled boiling for rectangular channels. Measurements of pressure drop on rectangular channel 60 and 90 cm long and with a coolant gap of 1,8 and 3,6 mm were performed in the following range of parameters. -) 3 < pressure at the outlet < 11 bars abs; -) 25 < inlet temperature < 70 deg. C; -) 200 < heat flux < 700 W/cm 2 . It appeared that the usual parameter, relative length in subcooled boiling, was not sufficient to correlate experimental pressure losses on the subcooled boiling length and that there was a supplementary influence of pressure, heat flux and subcooling. With an a dimensional parameter including these terms a correlation was established with an error band of ±10%. With a computer code it was possible to derive the relation giving the overall pressure drop along the channel and to determine the local gradients of pressure drop. These local gradients were then correlated with the above parameter calculated in local conditions. 95 % of the experimental points were computed with an accuracy of ±10% with this correlation of gradients which can be used for non-uniform heated channels. (authors) [fr

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

OpenAIRE

Hwang Bae; Dong Eok Kim; Sung-Uk Ryu; Sung-Jae Yi; Hyun-Sik Park

2017-01-01

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are s...

Investigation of small break loss-of-coolant phenomena in a small scale nonnuclear test facility

International Nuclear Information System (INIS)

Cozzuol, J.M.; Fauble, T.J.; Harvego, E.A.

1980-01-01

A small-scale nonnuclear integral test facility designed to simulate a pressurized water reactor (PWR) system was used to evaluate the effects of a small break loss-of-coolant accident (LOCA) on the system thermal-hydraulic response. The experiment approximated a 2.5% (11-cm diameter) communicative break in the cold leg of a PWR, and included initial conditions which were similar to conditions in a PWR operating at full power. The 2.5% break size ensured that the nominal break flow rate was greater than the high pressure injection system (HPIS) flow rate, thus providing the potential for a continuous system depressurization. The sequence of events was similar to that used in evaluation model analysis of small break loss-of-coolant accidents, and included simulated reactor scram and loss of offsite power. Comparisions of experimental data with computer code calculations are used to demonstrate the capabilities and limitations of integral system calculations used to predict phenomena which can be important in the assessment of a small break LOCA in a PWR

Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

Energy Technology Data Exchange (ETDEWEB)

Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

2016-11-15

The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

Continuous positive airway pressure breathing increases cranial spread of sensory blockade after cervicothoracic epidural injection of lidocaine.

NARCIS (Netherlands)

Visser, W.A.; Eerd, M.J. van; Seventer, R. van; Gielen, M.J.M.; Giele, J.L.P.; Scheffer, G.J.

2007-01-01

BACKGROUND: Continuous positive airway pressure (CPAP) increases the caudad spread of sensory blockade after low-thoracic epidural injection of lidocaine. We hypothesized that CPAP would increase cephalad spread of blockade after cervicothoracic epidural injection. METHODS: Twenty patients with an

Failure rates in Barsebaeck-1 reactor coolant pressure boundary piping. An application of a piping failure database

Energy Technology Data Exchange (ETDEWEB)

Lydell, B. [RSA Technologies, Vista, CA (United States)

1999-05-01

This report documents an application of a piping failure database to estimate the frequency of leak and rupture in reactor coolant pressure boundary piping. The study used Barsebaeck-1 as reference plant. The study tried two different approaches to piping failure rate estimation: 1) PSA-style, simple estimation using Bayesian statistics, and 2) fitting of statistical distribution to failure data. A large, validated database on piping failures (like the SKI-PIPE database) supports both approaches. In addition to documenting leak and rupture frequencies, the SKI report describes the use of piping failure data to estimate frequency of medium and large loss of coolant accidents (LOCAs). This application study was co sponsored by Barsebaeck Kraft AB and SKI Research 41 refs, figs, tabs

An evaluation of debris mobility within a PWR reactor coolant system during the recirculation mode

International Nuclear Information System (INIS)

Andreychek, T.S.

1987-01-01

To provide for the long-term cooling of the nuclear core of a Pressurized Water Rector (PWR) following a hypothetical Loss-of-Coolant Accidnet (LOCA), water is drawn from the containment sump and pumped into the reactor coolant system (RCS). It has been postulated that debris from the containment, such as dirt, sand, and paint from containment walls and in-containment equipment, could be carried into the containment sump due to the action of the RCS coolant that escapes from the breach in the piping and then flows to the sump. Once in the sump, this debris could be pumped into the Safety Injection System (SIS) and ultimately the RCS itself, causing the performance of the SIS to be degraded. Of particular interest is the potential for core blockage that may occur due to debris transport into the core region by the recirculating flow. This paper presents a method of evaluating the potential for debris from the sump to form core blockages under recirculating flow conditions following a hypothetical LOCA for a PWR

Coolant mixing in pressurized water reactors. Proceedings

International Nuclear Information System (INIS)

Hoehne, T.; Grunwald, G.; Rohde, U.

1998-10-01

For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

Integrated equipment for increasing and maintaining coolant pressure in primary circuit of PWR nuclear power plant

International Nuclear Information System (INIS)

Sykora, D.

1986-01-01

An open heat pump circuit is claimed connected to the primary circuit. The pump circuit consists of a steam pressurizer with a built-in steam distributor, a compressor, an expander, a reducing valve, an auxiliary pump, and of water and steam pipes. The operation is described and a block diagram is shown of integrated equipment for increasing and maintaining pressure in the nuclear power plant primary circuit. The appropriate entropy diagram is also shown. The advantage of the open pump circuit consists in reducing the electric power input and electric power consumption for the steam pressurizers, removing entropy loss in heat transfer with high temperature gradient, in the possibility of inserting, between the expander and the auxiliary pump, a primary circuit coolant treatment station, in simplified design and manufacture of the high-pressure steam pressurizer vessel, reducing the weight of the steam pressurizer by changing its shape from cylindrical to spherical, increasing the rate of pressure growth in the primary circuit. (E.S.)

Steam CFD simulation of injection in suppression pool

International Nuclear Information System (INIS)

Naveen Samad, A.M.; Ghosh, Sumana

2015-01-01

Boiling water reactor (BWR) is one of the common types of electricity generating nuclear reactor. Suppression pool system is a major component of the BWR which has to be designed efficiently for the safe operations. During some accidents like Loss of Coolant Accident (LOCA) large amount of steam are injected to the pressure suppression system resulting in increase in temperature of the pool and thereby increasing the pressure. The present work discuss about the Computational Fluid Dynamics (CFD) simulation of steam injected to the wet well of BWR through the blow down pipes and there by investigating the hydrodynamic and thermal characteristics of the system. The simulations were carried out for three different steam injection velocities. The numerical simulations were performed with ANSYS FLUENT using multiphase 3D Volume of Fluid (VOF) model and k-ε model was adopted for modelling turbulence flow. (author)

Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

Energy Technology Data Exchange (ETDEWEB)

Seeliger, André, E-mail: a.seeliger@hszg.de [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Alt, Sören; Kästner, Wolfgang; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Kryk, Holger; Harm, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany)

2016-08-15

Highlights: • Borated coolant supports corrosion at zinc-coated installations in PWR after LOCA. • Dissolved zinc is injected into core by ECCS during sump recirculation phase. • Corrosion products can reach and settle at further downstream components. • Corrosion products can cause head losses at spacers and influence decay heat removal. • Preventive procedures were tested at semi-technical scale facilities. - Abstract: Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized compounds, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively. First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the Hochschule Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators. As near-plant core components, a 3 × 3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module includes segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid

Analysis of Pressure Pulsation Induced by Rotor-Stator Interaction in Nuclear Reactor Coolant Pump

Directory of Open Access Journals (Sweden)

Xu Zhang

2017-01-01

Full Text Available The internal flow of reactor coolant pump (RCP is much more complex than the flow of a general mixed-flow pump due to high temperature, high pressure, and large flow rate. The pressure pulsation that is induced by rotor-stator interaction (RSI has significant effects on the performance of pump; therefore, it is necessary to figure out the distribution and propagation characteristics of pressure pulsation in the pump. The study uses CFD method to calculate the behavior of the flow. Results show that the amplitudes of pressure pulsation get the maximum between the rotor and stator, and the dissipation rate of pressure pulsation in impellers passage is larger than that in guide vanes passage. The behavior is associated with the frequency of pressure wave in different regions. The flow rate distribution is influenced by the operating conditions. The study finds that, at nominal flow, the flow rate distribution in guide vanes is relatively uniform and the pressure pulsation amplitude is the smallest. Besides, the vortex shedding or backflow from the impeller blade exit has the same frequency as pressure pulsation but there are phase differences, and it has been confirmed that the absolute value of phase differences reflects the vorticity intensity.

Effect of Fuel Injection and Mixing Characteristics on Pulse-Combustor Performance at High-Pressure

Science.gov (United States)

Yungster, Shaye; Paxson, Daniel E.; Perkins, Hugh D.

2014-01-01

Recent calculations of pulse-combustors operating at high-pressure conditions produced pressure gains significantly lower than those observed experimentally and computationally at atmospheric conditions. The factors limiting the pressure-gain at high-pressure conditions are identified, and the effects of fuel injection and air mixing characteristics on performance are investigated. New pulse-combustor configurations were developed, and the results show that by suitable changes to the combustor geometry, fuel injection scheme and valve dynamics the performance of the pulse-combustor operating at high-pressure conditions can be increased to levels comparable to those observed at atmospheric conditions. In addition, the new configurations can significantly reduce the levels of NOx emissions. One particular configuration resulted in extremely low levels of NO, producing an emission index much less than one, although at a lower pressure-gain. Calculations at representative cruise conditions demonstrated that pulse-combustors can achieve a high level of performance at such conditions.

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

Porous media heat transfer for injection molding

Science.gov (United States)

Beer, Neil Reginald

2016-05-31

The cooling of injection molded plastic is targeted. Coolant flows into a porous medium disposed within an injection molding component via a porous medium inlet. The porous medium is thermally coupled to a mold cavity configured to receive injected liquid plastic. The porous medium beneficially allows for an increased rate of heat transfer from the injected liquid plastic to the coolant and provides additional structural support over a hollow cooling well. When the temperature of the injected liquid plastic falls below a solidifying temperature threshold, the molded component is ejected and collected.

Zinc injection on the EDF pressurized light water reactors. Current results and operating experience feedback

International Nuclear Information System (INIS)

Piana, Olivier; Duval, Arnaud; Moleiro, Edgar; Benfarah, Moez; Bretelle, Jean-Luc; Chaigne, Guy

2014-01-01

Nowadays, zinc injection, as well as pH management and hydrogen control, is increasingly considered as an essential element of PWR Primary Water Chemistry worldwide. After a first implementation of zinc injection at Bugey 2 since 2004 and Bugey 4 since 2006, EDF decided to extend this practice, which constitutes a modification of primary circuit chemical conditioning, to other units of its fleet. Currently, 15 among the 58 reactors of the French fleet are injecting depleted zinc acetate into the primary coolant water. Three main goals were identified at the beginning of this program. Indeed, the expected benefits of zinc injection were: Reduction of the rate of generalized corrosion and mitigation of stress corrosion cracking initiation on nickel based alloys (Material goal). Curative or preventive reduction of radiation sources to which workers are exposed (Radiation fields' goal). Mitigation of the AOA or CIPS risks by reduction of corrosion products releases and mitigation of crud deposition (Fuel protection goal). To monitor the zinc addition, EDF has defined a complete survey program concerning: chemistry and radiochemistry responses (primary coolant monitoring of corrosion and fission products and calculation of zinc injected, zinc removed and zinc incorporated in RCS surfaces) ; radiation fields (dose rates and deposited activities measurements) ; materials (statistical analysis of SG tube cracks) ; fuel (oxide thickness measurements and visual exams) ; effluents (corrosion products releases and isotopic distribution follow up) ; wastes (radiochemical characterization of filters). This paper will detail the present results of this monitoring program. It appears that the expected benefits of zinc injection have yet to be fully realized; further operating experience will be required in order to fully evaluate its impact. (author)

LHI (low head safety injection) emergency cooling pump test for the EPR trademark in operation with solid matter loaded water

International Nuclear Information System (INIS)

Ganzmann, I.; Schulte, C.

2010-01-01

Emergency cooling pumps are essential and indispensable components of the NPP safety philosophy. In case of a loss-of coolant accident solid matter (debris: fibrous insulation material, concrete dust, pigment particles) might be released into the coolant, LHSI (low head safety injection) pumps have to ensure their performance capacity for a certain amount of debris without damage or loss of power. The authors describe the development of a test facility. The LHSI was tested in continuous operation over a time period of 14 days with a debris content of 1500 ppm (90% mineral wool fibers, 3% concrete dust, 3% pigment particles, 4% microporous insulation material). The pump did not show any damage or loss of hydraulic power. Further tests including thermoshock conditions (temperature changes of 160 C) are planned.

Vent clearing during a simulated loss-of-coolant accident in Mark I boiling-water-reactor pressure-suppression system

International Nuclear Information System (INIS)

Pitts, J.H.; McCauley, E.W.

1978-01-01

The response of the pressure-suspension containment system of Mark I boiling-water reactors to a loss-of-coolant accident (LOCA) is being studied. This response is a design basis for light-water nuclear reactors. Part of the study is being carried out on a 1 / 5 -scale experimental facility that models the pressure-suppression containment system of the Peach Bottom 2 nuclear power plant. The test series reported here focused on the initial or air-clearing phase of a hypothetical LOCA. Measured forces, measured pressures, and the hydrodynamic phenomena (observed with high-speed cameras) show a logical interrelationship

Pressurization of a compartment due to the rupture of coolant piping

International Nuclear Information System (INIS)

Kot, C.A.; Hsieh, B.J.

1993-01-01

The pressurization and venting of enclosed compartments due to the accidental rupture of coolant piping is a safety problem common to many nuclear facilities. The processes associated with such an accident are very complex, involving, in general, transient multiphase flows, interactions and mixing between the incoming flows and the gases in the compartment, and heat transfer with the surroundings. Since pipe rupture is associated with many phenomenological uncertainties, elaborate 3-D thermal-hydraulic modeling and extensive calculational efforts are not warranted for many design applications. It is then more appropriate to rely. on simplified, global analysis approaches which can provide reasonably conservative estimates of the structural loads and flow processes, and which can readily be used in parameter/design studies. The objective of this paper is to present such an approach

Analysis of an ultrasonic level device for in-core Pressurized Water Reactor coolant detection

International Nuclear Information System (INIS)

Johnson, K.R.

1981-01-01

A rigorous semi-empirical approach was undertaken to model the response of an ultrasonic level device (ULD) for application to in-core coolant detection in Pressurized Water Reactors (PWRs). An equation is derived for the torsional wave velocity v/sub t phi/ in the ULD. Existing data reduction techniques were analyzed and compared to results from use of the derived equation. Both methods yield liquid level measurements with errors of approx. 5%. A sensitivity study on probe performance at reactor conditions predicts reduced level responsivity from data at lower temperatures

Experimental investigation of the concomitant injection of gasoline and CNG in a turbocharged spark ignition engine

International Nuclear Information System (INIS)

Momeni Movahed, M.; Basirat Tabrizi, H.; Mirsalim, M.

2014-01-01

Highlights: • Concomitant injection of gasoline and CNG is compared with gasoline and CNG modes. • BSFC, HC and CO emissions of the concomitant injection are lower than gasoline mode. • Deteriorations of the concomitant injection are negligible compared to gasoline mode. • Cylinder peak pressure and heat loss to coolant of the concomitant injection are lower than CNG mode. • Some shortcomings in CNG mode can be solved by changing the spark timing and lambda. - Abstract: Concomitant injection of gasoline and CNG is a new concept to overcome problems of bi-fueled spark ignition engines, which operate in single fuel mode, either in gasoline or in CNG mode. This experimental study indicates how some problems of gasoline mode such as retarded ignition timings for knock prevention and rich air–fuel mixture for component protection can be resolved with the concomitant injection of gasoline and CNG. Results clearly show that the concomitant injection improves thermal efficiency compared to gasoline mode. On the other hand, simultaneous injection of gasoline and CNG reduces some problems of CNG mode such as high cylinder pressure and heat loss to the engine coolant. This decreases the stringent requirements for thermal and mechanical strength of the engine components in CNG mode. In addition, it is shown that by modifying the spark advance and air fuel ratio in CNG mode, the engine operation improves in terms of NOx emissions and maximum in-cylinder pressure as the concomitant injection does. Nevertheless, new requirements such as an intercooler with higher cooling capacity are implied to the engine configuration. Finally, the most important concerns in control strategies of the engine control unit for a vehicle with concomitant injection of gasoline and CNG are discussed

Investigation of break location effects on thermal-hydraulics during intermediate break loss-of-coolant accident experiments at ROSA-III

International Nuclear Information System (INIS)

Koizumi, Yasuo; Tasaka, Kanji

1986-01-01

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25 % main recirculation pump suction line break (MRPS-B) experiments, the 21 % single-ended jet pump drive line break (JPD-B) experiment and the 15 % main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests. In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop. (author)

Design of automotive engine coolant hoses

Directory of Open Access Journals (Sweden)

Hrishikesh D BACHCHHAV

2018-03-01

Full Text Available In this paper, we are present the performance of engine coolant hoses (radiator hoses used in passenger cars by checking various physical behaviours such as hose leakage, hose burst, hose collapse or any mechanical damage as studied-thru design guidelines, CFD analysis and product validation testing and also check pressure drop of the hoses when engine will be running. The design term is more likely used for technical part modelling using CAD tool. Later on, we will focus on the transformation of the part design to process design. The process design term is more likely used for "tooling design" for manufacturing of the product using CAD Tool. Then inlet hose carries coolant from engine to radiator inlet tank, then coolant circulated in radiator and passed through radiator outlet tank to water pump of engine with the help of outlet hose. After that …nding any leakage, Burst, damage or collapse of hose and pressure drop of the hose with the help of design checklist, CFD Analysis and product validation testing.

Reactor auxiliary cooling facility and coolant supplying method therefor

International Nuclear Information System (INIS)

Ando, Koji; Kinoshita, Shoichiro.

1996-01-01

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

Reactor auxiliary cooling facility and coolant supplying method therefor

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1996-06-07

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

Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

International Nuclear Information System (INIS)

Kanzleiter, T.F.

1976-01-01

For the design of an LWR containment one of the important conditions to be considered is the rapid rise of internal pressure and temperature caused by a loss-of-coolant accident (LOCA) of the primary cooling system. The phenomena occurring within a containment during a LOCA are currently investigated through experiments with a model containment. The experimental results are compared with the results of model calculations to improve the calculational methods. An experimental facility was built, consisting of a primary coolant circuit and a special model containment. The model containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m 3 and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross sections. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiments a PWR configuration with nine compartments has been installed. The model scales of the compartment volumes and the overflow areas are about 1 : 64 compared to the 1200 MW PWR plant Biblis A. (Auth.)

Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

Energy Technology Data Exchange (ETDEWEB)

Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

1995-09-01

The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

Energy Technology Data Exchange (ETDEWEB)

Mukherjee, S; Ghosh, J K [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.; Patel, R J [Bhabha Atomic Research Centre, Mumbai (India). Refuelling Technology Division

1994-12-31

Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs.

On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

International Nuclear Information System (INIS)

Mukherjee, S.; Ghosh, J.K.; Patel, R.J.

1994-01-01

Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs

Development of fast-burn combustion with elevated coolant temperatures for natural gas engines. Final report, May 1985-May 1990

Energy Technology Data Exchange (ETDEWEB)

Bruch, K.L.; Dennis, J.W.

1990-09-01

The overall objective of the work was to improve the state of the art in the gas fired spark ignited engine for use in a cogeneration system. Four characteristics were enhanced for cogeneration, namely, Low Pressure Gas Induction, Improved Shaft Thermal Efficiency, Low NOx Emissions, and Increased Jacket Coolant Temperature. Using Taguchi methods and statistical design of experiment methodologies, an engine design evolved that exhibited: The ability to run satisfactorily on supply gas pressure as low as 1.5 psig (goal: 1 psig); A brake specific fuel consumption as low as 6950 Btu/hp-hr (36.6% thermal efficiency) at 2 gm/hp-hr NOx (goal: 7000 acceptable, 6800 excellent with NOx no more than 2 gm/hp-hr); A jacket water coolant system (with oil cooler on the same circuit) temperature of 225 F (goal); and The ability to burn gas with Methane Number as low as 67 (goal).

The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

2001-03-01

In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary.

The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

International Nuclear Information System (INIS)

Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

2001-03-01

In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary

Fusion-reactor blanket and coolant material compatibility

International Nuclear Information System (INIS)

Jeppson, D.W.; Keough, R.F.

1981-01-01

Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

Method and apparatus for suppressing water-solid overpressurization of coolant in nuclear reactor power apparatus

International Nuclear Information System (INIS)

Aanstad, O.J.; Sklencar, A.M.

1983-01-01

A reactor-coolant relief valve is opened for increase in mass influx if the rate of change of coolant pressure exceeds a setpoint during a predetermined interval, if, during this interval, the coolant temperature is less than a setpoint and if the level of the fluid in the pressurizer is above a predetermined setpoint (water-solid state). (author)

Pressure tube type reactors

International Nuclear Information System (INIS)

Komada, Masaoki.

1981-01-01

Purpose: To increase the safety of pressure tube type reactors by providing an additional ECCS system to an ordinary ECCS system and injecting heavy water in the reactor core tank into pressure tubes upon fractures of the tubes. Constitution: Upon fractures of pressure tubes, reduction of the pressure in the fractured tubes to the atmospheric pressure in confirmed and the electromagnetic valve is operated to completely isolate the pressure tubes from the fractured portion. Then, the heavy water in the reactor core tank flows into and spontaneously recycles through the pressure tubes to cool the fuels in the tube to prevent their meltdown. By additionally providing the separate ECCS system to the ordinary ECCS system, fuels can be cooled upon loss of coolant accidents to improve the safety of the reactors. (Moriyama, K.)

Pumps for German pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Dernedde, R.

1984-01-01

The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

Thermal shock studies associated with injection of emergency core coolant in pressurized water reactors

International Nuclear Information System (INIS)

Cheverton, R.D.; Bolt, S.E.; Iskander, S.K.

1977-01-01

Studies to determine the accuracy of calculational techniques for predicting crack initiation and arrest in PWR vessels due to thermal shock from ECC injection are described. The reference calculational model is reviewed, the experimental program and facilities are described, and some thermal shock experiments and results are discussed

Light extinction method on high-pressure diesel injection

Science.gov (United States)

Su, Tzay-Fa; El-Beshbeeshy, Mahmound S.; Corradini, Michael L.; Farrell, Patrick V.

1995-09-01

A two dimensional optical diagnostic technique based on light extinction was improved and demonstrated in an investigation of diesel spray characteristics at high injection pressures. Traditional light extinction methods require the spray image to be perpendicular to the light path. In the improved light extinction scheme, a tilted spray image which has an angle with the light path is still capable of being processed. This technique utilizes high speed photography and digital image analysis to obtain qualitative and quantitative information of the spray characteristics. The injection system used was an electronically controlled common rail unit injector system with injection pressures up to 100 MPa. The nozzle of the injector was a mini-sac type with six holes on the nozzle tip. Two different injection angle nozzles, 125 degree(s) and 140 degree(s), producing an in-plane tilted spray and an out of plane tilted spray were investigated. The experiments were conducted on a constant volume spray chamber with the injector mounted tilted at an angle of 62.5 degree(s)$. Only one spray plume was viewed, and other sprays were free to inject to the chamber. The spray chamber was pressurized with argon and air under room temperature to match the combustion chamber density at the start of the injection. The experimental results show that the difference in the spray tip penetration length, spray angle, and overall average Sauter mean diameter is small between the in- plane tilted spray and the out of plane tilted spray. The results also show that in-plane tilted spray has a slightly larger axial cross- section Sauter mean diameter than the out of plane tilted spray.

Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

OpenAIRE

Lefèvre, Grégory; Živković, Ljiljana S.; Jaubertie, Anne

2012-01-01

In the primary circuit of pressurized water reactors (PWR), the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspec...

Precision tubes for high-pressure diesel injection lines; Praezisrohre fuer Hochdruck-Dieseleinspritzleitungen

Energy Technology Data Exchange (ETDEWEB)

Hagedorn, M.; Lechtenfeld, U.; Zaremba, A. [Mannesmann Praezisrohr GmbH, Hamm (Germany)

2008-03-15

The requirements on diesel injection lines raise because of increasing customers demands and more rigid environmental laws. In this context higher injection pressures effect both aspects positively. One important condition for increasing pressure levels is the economical provision of suitable injection lines. To reach this aim, Mannesmann Praezisrohr GmbH developed precision tubes for injection lines, which are fulfilling these increasing requirements. (orig.)

Flow rate control systems for coolants for BWR type reactors

International Nuclear Information System (INIS)

Igarashi, Yoko; Kato, Naoyoshi.

1981-01-01

Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

Long-term follow-up of high-pressure injection injuries to the hand

NARCIS (Netherlands)

Wieder, Anat; Lapid, Oren; Plakht, Ygal; Sagi, Amiram

2006-01-01

High-pressure injection injury is an injury caused by accidental injection of substances by industrial equipment. This injury may have devastating sequelae. The goal of this study was to assess the long-term outcome of high-pressure injection injury to the hand. In this historical prospective study,

Installation for low temperature vapor explosion experiment

International Nuclear Information System (INIS)

Nilsuwankosit, Sunchai; Archakositt, Urith

2000-01-01

A preparation for the experiment on the low temperature vapor explosion was planned at the department of Nuclear Technology, Chulalongkorn University, Thailand. The objective of the experiment was to simulate the interaction between the molten fuel and the volatile cooling liquid without resorting to the high temperature. The experiment was expected to involve the injection of the liquid material at a moderate temperature into the liquid material with the very low boiling temperature in order to observe the level of the pressurization as a function of the temperatures and masses of the applied materials. For this purpose, the liquid nitrogen and the water were chosen as the coolant and the injected material for this experiment. Due to the size of the installation and the scale of the interaction, only lumped effect of various parameters on the explosion was expected from the experiment at this initial stage. (author)

Small break LOCA [loss of coolant accident] mitigation for Bellefonte

International Nuclear Information System (INIS)

Bayless, P.D.; Dobbe, C.A.

1986-01-01

Several 5-cm (2-in.) diameter cold leg break loss coolant accidents for the Bellefonte nuclear plant were analyzed as part of the Severe Accident Sequence Analysis Program. The transients assumed various system failures, and included the S 2 D sequence. Operator actions to mitigate the S 2 D transient were also investigated. The transients were analyzed until either core damage began or long-term decay heat removal was established. The S 2 D sequence was analyzed into the core damage phase of the transient. The analyses showed that the flow from one high pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were also able to prevent core damage for the S 2 D sequence

LOFT/LP-SB-3, Loss of Fluid Test, Cold Leg Break LOCA, No High Pressure injection System (HPIS)

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The sixth OECD LOFT experiment was conducted on 5 March 1984. It simulated a 1.8-in cold leg break LOCA with no HPIS available. This experiment was designed mainly for investigation of plant recovery effectiveness using secondary bleed and feed during core uncover and addressed accumulator injection at low pressure differentials. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

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.)

THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

International Nuclear Information System (INIS)

Muramatsu, Ken; Akimoto, Masayuki

1982-08-01

THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

International Nuclear Information System (INIS)

Zhang Longfei; Zhang Dafa; Wang Shaoming

2007-01-01

Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

Data report for ROSA-IV LSTF gravity-driven safety injection experiment run SB-CL-27

International Nuclear Information System (INIS)

Yonomoto, Taisuke; Saitou, Seishi; Kuroda, Takeshi

1994-03-01

Experimental data are presented for the passive injection test, Run SB-CL-27, conducted at the ROSA-IV Large Scale Test Facility (LSTF) on September 17, 1992. This experiment simulated thermal-hydraulic behavior of a gravity-driven, passive safety injection system during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). The injection system consisted of a gravity-driven injection tank, located above the reactor vessel, with connecting lines. The tank was initially filled with water of room temperature at the same pressure as the pressurizer. The connecting lines to the cold leg and to the vessel downcomer were opened at the test initiation. Then, a natural circulation flow developed in the loop which was formed by these lines and the injection tank. The hot water in the cold leg circulated into the upper part of tank and accumulated there causing a significant thermal stratification. This thermal stratification prevented direct-contact condensation of steam from occurring during the subsequent tank drain-down phase. Therefore, no condensation-induced depressurization of the tank, affecting adversely the injection performance, occurred. (author)

Fast measurements of the in-core coolant velocity in a BWR by neutron noise analysis

International Nuclear Information System (INIS)

Hagen, T.H.J.J. van der; Hoogenboom, J.E.

1988-01-01

A method to determine in-core coolant velocities from neutron noise within short time intervals has been developed. The accuracy of the method was determined by using a simulation set-up and by using signals of a twin self-powered neutron detector installed in the core of the Dodewaard BWR in the Netherlands. In-core coolant velocities can be estimated within 2.5 s with a standard deviation (due to statistics) less than 2.1%. The method is suitable for velocity monitoring as is shown by the application to a stepwise velocity change of the coolant in a model of a coolant channel of a BWR. The presented technique was applied to determine the variations of the coolant velocity in the Dodewaard core during normal operation and during pressure steps. Only minor variations of the coolant velocity were detected during normal reactor conditions. An increase of those variations with pressure lowering - indicating a lower thermal hydraulic stability - could be detected. A clear velocity response to pressure steps could be determined which was also reflected in the cross-spectrum of the velocity with the vessel pressure and with the in-core neutron flux. (author)

CONTEMPT: computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

International Nuclear Information System (INIS)

Hsii, Y.H.

1978-04-01

The CONTEMPT code is used by Babcock and Wilcox for containment analysis following a postulated loss of coolant accident. An additional model is described which is used for the calculation of long term post reflood mass and energy releases to the containment that is used for the containment design basis LOCA calculations. These calculations maximize the rate of energy flow to the containment. The mass and energy data are given to the containment designer for use in calculating the containment building design pressure and temperature and in sizing containment heat removal equipment

Recent development in PWR zinc injection

International Nuclear Information System (INIS)

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

2002-01-01

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

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.)

Neutronic Analysis on Coolant Options in a Hybrid Reactor System for High Level Waste Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Seoul (Korea, Republic of)

2014-10-15

A fusion-fission hybrid reactor (FFHR) which is a combination of plasma fusion tokamak as a fast neutron source and a fission reactor as of fusion blanket is another potential candidate. In FFHR, fusion plasma machine can supply high neutron-rich and energetic 14.1MeV (D, T) neutrons compared to other options. Therefore it has better capability in HLW incineration. While, it has lower requirements compared to pure fusion. Much smaller-sized tokamak can be achievable in a near term because it needs relatively low plasma condition. FFHR has also higher safety potential than fast reactors just as ADSR because it is subcritical reactor system. FFHR proposed up to this time has many design concepts depending on the design purpose. FFHR may also satisfy many design requirement such as energy multiplication, tritium production, radiation shielding for magnets, fissile breeding for self-sustain ability also waste transmutation. Many types of fuel compositions and coolant options have been studied. Effect of choices for fuel and coolant was studied for the transmutation purpose FFHR by our team. In this study LiPb coolant was better than pure Li coolant both for neutron multiplication and tritium breeding. However, performance of waste transmutation was reduced with increased neutron absorption at coolant caused by tritium breeding. Also, LiPb as metal coolant has a problem of massive MHD pressure drop in coolant channels. Therefore, in a previous study, waste transmutation performance was evaluated with light water coolant option which may be a realistic choice. In this study, a neutronic analysis was done for the various coolant options with a detailed computation. One of solutions suggested is to use the pressure tubes inside of first wall and second wall In this work, performance of radioactive waste transmutation was compared with various coolant options. On the whole, keff increases with all coolants except for FLiBe, therefore required fusion power is decreased. In

Fundamental study of a water jet injected into a vacuum vessel of fusion reactor under the ingress of coolant event

International Nuclear Information System (INIS)

Takase, Kazuyuki; Kunugi, Tomoaki; Seki, Yasushi; Kurihara, Ryouichi; Ueda, Shuzou

1996-01-01

As one of some transient sequences for the thermofluid safety in ITER, pressure rise and boiling heat transfer characteristics in a Tokamak vacuum vessel during an ingress of coolant event (ICE) are being investigated experimentally by using the preliminary ICE apparatus. The pressure rise rates in the vacuum vessel and the wall temperature distributions on the target plate were measured quantitatively and clarified at first. In addition, a two-phase flow under the ICE conditions was analyzed numerically for predicting the experimental results using one-dimensional transport equations and the drift-flux model. The experimental results were compared with the numerical results. It was found that the pressurization behavior during the ICE conditions could be estimated qualitatively by the present numerical analyses. 5 refs., 5 figs

Bandwidth of reactor internals vibration resonance with coolant pressure oscillations

International Nuclear Information System (INIS)

Proskuryakov, K.N.; Novikov, K.S.; Galivec, E.Yu.

2009-01-01

In a few decades a significant increase in a part of an electricity development on the NPP will require NPP to be operated in non full capacity modes and increase in operation time in transitive modes. Operating in such conditions as compared to the operation on a constant mode will lead to the increase in cyclic dynamical loading. In water cooled water moderated reactors these loading are realized as low-cyclic and high-cyclic loadings. High-cyclic loadings increases are caused by a raised vibration in non stationary modes of operation. It is known, that in some modes of a non full capacity reactor high-cyclic dynamic loadings can increase. It is obvious, that the development of management technologies is necessary for the life time management operation. In the context of this problem one of the main tasks are revealing and the prevention of the conditions of the occurrence of the operation leading to the resonant interaction of the coolant fluctuations and the equipment, reactor vessel (RV), fuel assemblies (FA) and reactor internals (RI) vibration. To prevent the appearance of the conditions for resonance interaction between the fluid flow and the equipments, it is necessary to provide the different frequencies for the self oscillations in the separated elements of the circulating system and also in the parts of the system formed by the comprising of these elements. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of coolant outside of which there is no resonant interaction. The presented work is devoted to finding the solution of this problem. There are results of theoretical an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. The accordance of results had been calculated with had been measured are satisfied for practical purposes. These

Organic coolant in Winnipeg riverbed sediments

International Nuclear Information System (INIS)

Guthrie, J.E.; Acres, O.E.

1979-03-01

Between January and May 1977 a prolonged leak of organic coolant occurred from the Whiteshell Nuclear Research Establishment's nuclear reactor, and a minimum of 1450 kg of coolant entered the Winnipeg River and was deposited on the riverbed. The level of radioactivity associated with this coolant was low, contributing less than 0.2 μGy (0.02 mrad) a year to the natural background gamma radiation field from the riverbed. The concentration of coolant in the water samples never exceeded 0.02 mg/L, the lower limit of detection. The mortality of crayfish, held in cages where the riverbed was covered with the largest deposits of coolant, was not significantly different from that in the control cages upstream of the outfall. No evidence of fish kill was found. (author)

A simulation experiment and analysis on the effects of in-coherence in fuel coolant interaction

International Nuclear Information System (INIS)

Kondo, S.; Togo, Y.; Iwamura, T.

1976-01-01

Experimental and analytical studies were conducted to investigate effects of incoherence (space time behavior of molten fuel) on molten fuel coolant interaction. In experiments, a 2 mm diameter molten tin jet was injected upward into the water in a slender tank. The results were analyzed based on the pressure records and high speed photographs. The pressure records indicated that there were two types of interaction between molten jet and water, intermittent explosion mode and continuous one. The explosion mode appeared when the temperature of molten tin was above 350 0 C or so and that of water was below 70 0 C or so. The high speed photograph indicated that an establishment of a stable jet column was necessary for an explosive interaction and that a bubble like region grew and collapsed at the root of the jet in accordance with the generation of pressure pulse. It was found that the mass of metal which contributed to the vapor explosion was only a small part of the injected metal in the case of jet injection type contact mode and this was the reason why the gross thermal to mechanical energy conversion ratio was around 0.03% in this type of contact mode, though this ratio was around 2% if only the part of record around the pressure pulse was taken into consideration. In the analysis part, a multi-channel FCI model was developed to evaluate the spatial incoherence effect on pressure at subassembly exit. The calculated pressure trace indicated that the spatial incoherence has considerable effects for an evaluation of structure response under FCI pressure loading. (auth.)

Design of Safety Injection Tanks Using Axiomatic Design and TRIZ

International Nuclear Information System (INIS)

Heo, Gyunyoung; Jeong, Yong Hoon

2008-01-01

Design can be categorized into two steps: 'synthesis' and 'analysis'. While synthesis is the process of decision-making on design parameters, analysis is the process of optimizing the parameters selected. It is known from experience that the mistakes made in the synthesis process are hardly corrected in the analysis process. 'Systematic synthesis' is, therefore, easy to overlook but an important topic. 'Systematic' is interpreted as 'minimizing' uncertainty and subjectivity. This paper will introduce the design product achieved by using Axiomatic Design (AD) and TRIZ (Theory of Inventive Problem Solving romanized acronym for Russian), which is a new design of Safety Injection Tank (SIT). In designing a large-capacity SIT which should play an important role in mitigating the large break loss of coolant accidents, there are three issues: 1) the excessively large plenum for pressurized nitrogen gas; 2) the difficulties maintaining the high initial injection flow rate; and 3) the non-condensable nitrogen gas in the coolant. This study proposes a conceptual idea for SITs that are pressurized by the chemical reaction of solid propellants. The AD theory and the principles of TRIZ enable new approach in problem-solving for those three issues in an innovative way. The paper made an effort to clarify the systematic synthesis process to reach the final design solution. (authors)

Attempt of lean burn of a 4 cycle gasoline engine by the aid of low pressure air assisted in-cylinder injection; Tonai kuki nenryo funsha ni yoru lean burn no kokoromi

Energy Technology Data Exchange (ETDEWEB)

Hatakeyama, S; Kondo, M; Sekiya, Y; Murayama, T [Hokkaido Automotive Engineering College, Hokkaido (Japan)

1997-10-01

Comparable performance and exhaust emission with conventional carburetor was obtained by a low Pressure air assisted in-cylinder injection system. And lean burn of idling and light load operation till A/F=70 was realized by installing a spark Plug and a reed type injection nozzle in a divided combustion chambaer of a 4 cycle gasoline engine. 2 refs., 10 figs.

Compressed air injection technique to standardize block injection pressures : [La technique d'injection d'air comprimé pour normaliser les pressions d'injection d'un blocage nerveux].

Science.gov (United States)

Tsui, Ban C H; Li, Lisa X Y; Pillay, Jennifer J

2006-11-01

Presently, no standardized technique exists to monitor injection pressures during peripheral nerve blocks. Our objective was to determine if a compressed air injection technique, using an in vitro model based on Boyle's law and typical regional anesthesia equipment, could consistently maintain injection pressures below a 1293 mmHg level associated with clinically significant nerve injury. Injection pressures for 20 and 30 mL syringes with various needle sizes ( 18G, 20G, 21 G, 22G, and 24G) were measured in a closed system. A set volume of air was aspirated into a saline-filled syringe and then compressed and maintained at various percentages while pressure was measured. The needle was inserted into the injection port of a pressure sensor, which had attached extension tubing with an injection plug clamped "off". Using linear regression with all data points, the pressure value and 99% confidence interval (CI) at 50% air compression was estimated. The linearity of Boyle's law was demonstrated with a high correlation, r = 0.99, and a slope of 0.984 (99% CI: 0.967-1.001). The net pressure generated at 50% compression was estimated as 744.8 mmHg, with the 99% CI between 729.6 and 760.0 mmHg. The various syringe/needle combinations had similar results. By creating and maintaining syringe air compression at 50% or less, injection pressures will be substantially below the 1293 mmHg threshold considered to be an associated risk factor for clinically significant nerve injury. This technique may allow simple, real-time and objective monitoring during local anesthetic injections while inherently reducing injection speed. Présentement, aucune technique normalisée ne permet de vérifier les pressions d'injection pendant les blocages nerveux périphériques. Nous voulions vérifier si une technique d'injection d'air comprimé, utilisant un modèle in vitro fondé sur la loi de Boyle et du matériel propre à l'anesthésie régionale, pouvait maintenir avec régularité les

Experimental investigation of coolant and poisoned moderator mixing due to a simulated pressure tube/calandria tube fishmouth rupturing an overpoisoned guaranteed shutdown state

International Nuclear Information System (INIS)

Mackinnon, J.C.; Fortman, R.A.; Hadaller, G.I.

1997-01-01

During a guaranteed shutdown state (GSS) in a CANDU reactor, there must be sufficient negative reactivity to ensure subcriticality in the event of a process failure. In one of the acceptable states, the reactor is kept subcritical by a high concentration of a neutron-absorbing chemical (the poison gadolinium nitrate) dissolved in the moderator (i.e., the moderator is guaranteed overpoisoned). A postulated accident scenario which is considered as a part of reactor safety analysis is the rupture of a fuel channel (i.e., a pressure tube/calandria tube break) when the reactor is in a GSS. If one of the channels in the core breaks (requiring a simultaneous failure of both the pressure tube and the surrounding calandria tube), coolant from the primary heat transport system will be discharged into the moderator, causing an associated displacement of fluid through relief ducts at the top of the calandria vessel. The incoming (unpoisoned) coolant may mix quickly with the moderator, or may mix slowly while displacing poisoned moderator through the relief ducts. The effectiveness of mixing generally depends on the break location, the coolant discharge rate and the moderator circulation. If an in-core loss of coolant accident occurred while the reactor is in this overpoisoned state, it must be guaranteed that even with the dilution of the poison by the incoming coolant the reactor will remain subcritical on both a local and global basis. This paper presents an overview of an experimental program in progress at the Moderator Test Facility at Stern Laboratories to investigate coolant/poison mixing for a simulated in-core fishmouth pressure tube/calandria tube rupture. The nominal system at the same temperature as the heavily poisoned moderator, i.e., a depressurised 'cold' state. The results presented are those obtained during the commissioning of the modified Test Facility. The contents of the paper are as follows. First, the objectives of the experimental program are

Pneumomediastinum following high pressure air injection to the hand.

LENUS (Irish Health Repository)

Kennedy, J

2010-04-01

We present the case of a patient who developed pneumomediastinum after high pressure air injection to the hand. To our knowledge this is the first reported case of pneumomediastinum where the gas injection site was the thenar eminence. Fortunately the patient recovered with conservative management.

Pneumomediastinum following high pressure air injection to the hand.

LENUS (Irish Health Repository)

Kennedy, J

2012-02-01

We present the case of a patient who developed pneumomediastinum after high pressure air injection to the hand. To our knowledge this is the first reported case of pneumomediastinum where the gas injection site was the thenar eminence. Fortunately the patient recovered with conservative management.

Intraocular Pressure Increases After Intraarticular Knee Injection With Triamcinolone but Not Hyaluronic Acid.

Science.gov (United States)

Taliaferro, Kevin; Crawford, Alexander; Jabara, Justin; Lynch, Jonathan; Jung, Edward; Zvirbulis, Raimonds; Banka, Trevor

2018-03-09

Intraarticular steroid injections are a common first-line therapy for severe osteoarthritis, which affects an estimated 27 million people in the United States. Although topical, oral, intranasal, and inhalational steroids are known to increase intraocular pressure in some patients, the effect of intraarticular steroid injections on intraocular pressure has not been investigated, to the best of our knowledge. If elevated intraocular pressure is sustained for long periods of time or is of sufficient magnitude acutely, permanent loss of the visual field can occur. How does intraocular pressure change 1 week after an intraarticular knee injection either with triamcinolone acetonide or hyaluronic acid? A nonrandomized, nonblinded prospective cohort study was conducted at an outpatient, ambulatory orthopaedic clinic. This study compared intraocular pressure elevation before and 1 week after intraarticular knee injection of triamcinolone acetonide versus hyaluronic acid for management of primary osteoarthritis of the knee. Patients self-selected to be injected in their knee with either triamcinolone acetonide or hyaluronic acid before being informed of the study. The primary endpoint was intraocular pressure elevation of ≥ 7 mm Hg 1 week after injection. This cutoff is determined as the minimum significant pressure change in the ophthalmology literature recognized as an intermediate responder to steroids. Intraocular pressure was measured using a handheld Tono-Pen® applanation device. This device is frequently used in intraocular pressure measurement in clinical and research settings; 10 sequential measurements are obtained and averaged with a confidence interval. Only measurements with a 95% confidence interval were used. Over a 6-month period, a total of 96 patients were approached to enroll in the study. Sixty-two patients out of 96 approached (65%) agreed. Thirty-one (50%) were injected with triamcinolone and 31 (50%) were injected with hyaluronic acid. Patients

Analysis of accidental loss of pool coolant due to leakage in a PWR SFP

International Nuclear Information System (INIS)

Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

2015-01-01

Highlights: • Accidental loss of pool coolant due to leakage in a PWR SFP was studied using MAAP5. • The effect of emergency ventilation on the accident progression was investigated. • The effect of emergency injection on the accident progression was discussed. - Abstract: A large loss of pool coolant/water accident may be caused by extreme accidents such as the pool wall or bottom floor punctures due to a large aircraft strike. The safety of SFP under this circumstance is very important. Large amounts of radioactive materials would be easily released into the environment if a severe accident happened in the SFP, because the spent fuel pool (SFP) in a PWR nuclear power station (NPS) is often located in the fuel handing building outside the reactor containment. To gain insight into the loss of pool coolant accident progression for a pressurized water reactor (PWR) SFP, a computational model was established by using the Modular Accident Analysis Program (MAAP5). Important factors such as Zr oxidation by air, air natural circulation and thermal radiation were considered for partial and complete drainage accidents without mitigation measures. The calculation indicated that even if the residual water level was in the active fuel region, there was a chance to effectively remove the decay heat through axial heat conduction (if the pool cooling system failed) or steam cooling (if the pool cooling system was working). For sensitivity study, the effects of emergency ventilation and water injection on the accident progression were analyzed. The analysis showed that for the current configuration of high-density storage racks, it was difficult to cool the spent fuels by air natural circulation. Enlarging the space between the adjacent assemblies was a way of increasing air natural circulation flow rate and maintaining the coolability of SFP. Water injection to the bottom of the SFP helped to recover water inventory, quenching the high temperature assemblies to prevent

A Comparative Study on Energy and Exergy Analyses of a CI Engine Performed with Different Multiple Injection Strategies at Part Load: Effect of Injection Pressure

Directory of Open Access Journals (Sweden)

Muammer Özkan

2015-01-01

Full Text Available In this study, a four stroke four cylinder direct injection CI engine was run using three different injection pressures. In all measurements, the fuel quantity per cycle, the pre injection and main injection timing, the boost pressure and the engine speed were kept constant. The motor tests were performed under 130, 140 and 150 MPa rail pressure. During the theoretical part of the study, combustion, emission, energy and exergy analysis were made using the test results. An increase in the injection pressure increases combustion efficiency. The results show that combustion efficiency is not enough by itself, because the increase in the power need of the injection pump, decreases the thermal efficiency. The increase in the combustion temperature, increases the cooling loss and decreases the exergetic efficiency. In addition, the NOx emissions increased by 12% and soot emissions decreased 44% via increasing injection pressure by 17%. The thermal and exergetic efficiencies are found inversely proportional with injection pressure. Exergy destruction is found independent of the injection pressure and its value is obtained as ~6%.

Vent clearing during a simulated loss-of-coolant accident in a Mark I boiling-water reactor pressure-suppression system

International Nuclear Information System (INIS)

Pitts, J.H.; McCauley, E.W.

1978-01-01

In this test series, drywell pressurization rate, drywell overpressure, downcomer submergence, and overall vent system loss coefficient were varied to quantify the primary load sensitivities in the pressure suppression system. Extensive tests were conducted on a unique three-dimensional 1/5 scale model of the pressure suppression system a MARK-I BWR. They were focused on the initial or air cleaning phase of a hypothetical loss of coolant accident. As a result of the complete measurement system employed including multiple high speed cameras, the logical interrelationship between measured forces, measured pressures, and the hydrodynamic phenomena observed in high speed photographic pictures were established. The quantitative values from the 1/5 scale experiments can be applied to full scale plants using established scaling laws. (author)

Water vapor as a perspective coolant for fast reactors

International Nuclear Information System (INIS)

Kalafati, D.D.; Petrov, S.I.

1978-01-01

Based on analysis of foreign projects of nuclear power plants with steam-cooled fast reactors, it is shown that low breeding ratio and large doubling time were caused by using nickel alloys, high vapor pressure and small volume heat release. The possibility is shown of obtaining doubling time in the necessary limits of T 2 =10-12 years when the above reasons for steam-cooled reactors are eliminated. Favourable combination of thermophysical and thermodynamic properties of water vapor makes it perspective coolant for power fast reactors

Experimental study on the performance of the vapor injection refrigeration system with an economizer for intermediate pressures

Science.gov (United States)

Moon, Chang-Uk; Choi, Kwang-Hwan; Yoon, Jung-In; Kim, Young-Bok; Son, Chang-Hyo; Ha, Soo-Jung; Jeon, Min-Ju; An, Sang-Young; Lee, Joon-Hyuk

2018-04-01

In this study, to investigate the performance characteristics of vapor injection refrigeration system with an economizer at an intermediate pressure, the vapor injection refrigeration system was analyzed under various experiment conditions. As a result, the optimum design data of the vapor injection refrigeration system with an economizer were obtained. The findings from this study can be summarized as follows. The mass flow rate through the compressor increases with intermediate pressure. The compression power input showed an increasing trend under all the test conditions. The evaporation capacity increased and then decreased at the intermediate pressure, and as such, it became maximum at the given intermediate pressure. The increased mass flow rate of the by-passed refrigerant enhanced the evaporation capacity at the low medium pressure range, but the increased saturation temperature limited the subcooling degree of the liquid refrigerant after the application of the economizer when the intermediate pressure kept rising, and degenerated the evaporation capacity. The coefficient of performance (COP) increased and then decreased with respect to the intermediate pressures under all the experiment conditions. Nevertheless, there was an optimum intermediate pressure for the maximum COP under each experiment condition. Therefore, the optimum intermediate pressure in this study was found at -99.08 kPa, which is the theoretical standard medium pressure under all the test conditions.

Disruption mitigation with high-pressure helium gas injection on EAST tokamak

Science.gov (United States)

Chen, D. L.; Shen, B.; Granetz, R. S.; Qian, J. P.; Zhuang, H. D.; Zeng, L.; Duan, Y.; Shi, T.; Wang, H.; Sun, Y.; Xiao, B. J.

2018-03-01

High pressure noble gas injection is a promising technique to mitigate the effect of disruptions in tokamaks. In this paper, results of mitigation experiments with low-Z massive gas injection (helium) on the EAST tokamak are reported. A fast valve has been developed and successfully implemented on EAST, with valve response time  ⩽150 μs, capable of injecting up to 7 × 1022 particles, corresponding to 300 times the plasma inventory. Different amounts of helium gas were injected into stable plasmas in the preliminary experiments. It is seen that a small amount of helium gas (N_He≃ N_plasma ) can not terminate a discharge, but can trigger MHD activity. Injection of 40 times the plasma inventory impurity (N_He≃ 40× N_plasma ) can effectively radiate away part of the thermal energy and make the electron density increase rapidly. The mitigation result is that the current quench time and vertical displacement can both be reduced significantly, without resulting in significantly higher loop voltage. This also reduces the risk of runaway electron generation. As the amount of injected impurity gas increases, the gas penetration time decreases slowly and asymptotes to (˜7 ms). In addition, the impurity gas jet has also been injected into VDEs, which are more challenging to mitigate that stable plasmas.

Upgradation of design features of primary coolant pumps of Indian 220 MWe PHWR

International Nuclear Information System (INIS)

Sharma, S.S.; Mhetre, S.G.; Manna, M.M.

1994-01-01

Evolution in the design features of Primary Coolant Pump (PCP) had started in fifties for catering to stringent specification requirements of reactor coolant systems of larger capacity reactors of various kinds. Primary coolant pumps of PWR and PHWR are employed for circulating radioactive, pressurized hot water in a circuit consisting of reactor (heat source) and steam generator (heat sink). As primary coolant pump capacity decides the station capacity, larger capacity primary coolant pumps have been evolved. Since primary coolant pump pressure containing parts are part of Primary Heat Transport system envelope, the parts are designed, manufactured, inspected and tested in accordance with the applicable system guidelines. Flywheel is mounted on the motor shaft for increasing mass moment of inertia of pump motor rotor to meet the coast down requirements of reactor cooling system under Class-IV electrical power supply failure. Due to limited accessibility of the PCP (PCP installed in shut down accessible area), quick maintenance, condition monitoring, reliable shaft seal system/bearing system aspects have been of great concern to reactor owners and pump manufacturers. In this paper upgradation of design features of RAPS, MAPS and NAPS primary coolant pumps have been covered. (author). 4 figs., 1 tab

Study of the Effect of Clay Particles on Low Salinity Water Injection in Sandstone Reservoirs

Directory of Open Access Journals (Sweden)

Sina Rezaei Gomari

2017-03-01

Full Text Available The need for optimal recovery of crude oil from sandstone and carbonate reservoirs around the world has never been greater for the petroleum industry. Water-flooding has been applied to the supplement primary depletion process or as a separate secondary recovery method. Low salinity water injection is a relatively new method that involves injecting low salinity brines at high pressure similar to conventional water-flooding techniques, in order to recover crude oil. The effectiveness of low salinity water injection in sandstone reservoirs depends on a number of parameters such as reservoir temperature, pressure, type of clay particle and salinity of injected brine. Clay particles present on reservoir rock surfaces adsorb polar components of oil and modify wettability of sandstone rocks to the oil-wet state, which is accountable for the reduced recovery rates by conventional water-flooding. The extent of wettability alteration caused by three low salinity brines on oil-wet sandstone samples containing varying clay content (15% or 30% and type of clay (kaolinite/montmorillonite were analyzed in the laboratory experiment. Contact angles of mica powder and clay mixture (kaolinite/montmorillonite modified with crude oil were measured before and after injection with three low salinity sodium chloride brines. The effect of temperature was also analyzed for each sample. The results of the experiment indicate that samples with kaolinite clay tend to produce higher contact angles than samples with montmorillonite clay when modified with crude oil. The highest degree or extent of wettability alteration from oil-wet to intermediate-wet state upon injection with low salinity brines was observed for samples injected with brine having salinity concentration of 2000 ppm. The increase in temperature tends to produce contact angles values lying in the higher end of the intermediate-wet range (75°–115° for samples treated at 50 °C, while their corresponding

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)

Nonlinear dynamic analysis of nuclear reactor primary coolant systems

International Nuclear Information System (INIS)

Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

1979-01-01

The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

Composition and concentration of soluble and particulate matter in the coolant of the reactor primary cooling system of the Embalse nuclear power plant

International Nuclear Information System (INIS)

Chocron, Mauricio; Garcia Rodenas, Luis; La Gamma, Ana M.; Villegas, Marina; Fernandez, Alberto N.; Allemandi, Walter; Manera, Raul; Rosales, Hugo

2000-01-01

Nuclear power plants type PWR and PHWR (pressurized water reactor and pressurized heavy water reactor) have three coolant circuits which only exchange energy among them. The primary circuit, whose coolant extracts the reactor energy, the secondary circuit or water-steam cycle and the tertiary circuit which could be lake, river or sea water. The chemistry of the primary and secondary coolants is carefully controlled with the aim of minimizing the corrosion of structural materials. However, very low rates of corrosion are inevitable and one of the consequences of the corrosion processes is the presence of soluble and particulate matter in the coolant from where several problems associated with mass transfer arisen. In this way radioactive nuclides are transported out of the core to the steam generators, hydraulic resistance increases and heat transfer capability degrades. In the present paper some alternative techniques are proposed for the quantification of both, the particulate and soluble matter present in the coolant and their correspondent composition. Some results are also included and discussed. (author)

Development of natural convection heat transfer correlation for liquid metal with overlying boiling coolant

International Nuclear Information System (INIS)

Cho, Jae Seon; Suh, Kune Y.; Chung, Chang Hyun; Park, Rae Joon; Kim, Sang Baik

1999-01-01

Experimental study was performed to investigate the natural convection heat transfer characteristics and the crust formation of the molten metal pool concurrent with forced convective boiling of the overlying coolant. Tests were performed under the condition of the bottom surface heating in the test section and the forced convection of the coolant being injected onto the molten metal pool. The constant temperature and constant heater input power conditions were adopted for the bottom heating. Test results showed that the temperature distribution and crust layer thickness in the metal layer are appreciably affected by the heated bottom surface temperature of the test section, but not much by the coolant injection rate. The relationship between the Nu number and Ra number in the molten metal pool region is determined and compared with the correlations in the literature, and the experiment without coolant boiling. A new correlation on the relationship between the Nu number and Ra number in the molten metal pool with crust formation is developed from the experimental data

Study on a corrosion products behavior with zinc injection in PWR primary circuit based on the crud investigation in actual plant

International Nuclear Information System (INIS)

Hisamune Kenji; Nambu Tohru; Akutagawa Daisuke; Nagata Nobuaki; Shimizu, Yuichi; Nagamine, Kunitaka; Kogawa Noritaka

2012-09-01

Zinc injection into the reactor coolant of Pressurized Water Reactor (PWR) is one of the most effective techniques for removing radioactive cobalt from the chromite type spinel oxides in the inner layer of reactor coolant component surfaces. Many documents have reported that zinc injection applied plants have generally achieved positive dose rate reduction effects for the reactor coolant components and piping. Tsuruga Nuclear Power Plant unit 2 (hereafter Tsuruga unit 2) which applied zinc injection in 2005 has also achieved good dose reduction effects in most primary coolant piping. However, some piping without dose-rate reduction effects exist, it has come to an issue to plan radiation exposure reduction measures for the works around relevant piping. To plan a reasonable dose-rate reduction measures for respective components and piping in primary systems, it is considered necessary to understand the dose-rate reduction effects of zinc injection mechanically. Therefore, the surface oxides/deposits on piping and fuel cladding tubes with operational histories under zinc injection were investigated. As a result of the investigation, the changes in specific aspects of oxide layers and CRUDs are found as follow: - Owing to long-term zinc injection, deposits and oxides on the primary coolant piping are extremely reduced, the boundary between the outer oxide layer and the inner oxide layer become unclear with the thinning of the outer oxide layers. - At a Pressurizer Spray Piping, the dose-rate increases than before zinc injection. The oxide layers have the same 3-layer construction as before zinc injection, and the surface concentration of the radioactive cobalt is high. - At the fuel cladding tube surfaces, what is observed is the amount of metal nickel decrease and the chromium composition increase in the spinel type oxide From these results, it is suggested that, the oxide layers composed principally of zinc chromite spinel grow up on the pipe surface with zinc

Effect of fuel injection pressure and injection timing of Karanja biodiesel blends on fuel spray, engine performance, emissions and combustion characteristics

International Nuclear Information System (INIS)

Agarwal, Avinash Kumar; Dhar, Atul; Gupta, Jai Gopal; Kim, Woong Il; Choi, Kibong; Lee, Chang Sik; Park, Sungwook

2015-01-01

Highlights: • Effect of FIP on microscopic spray characteristics. • Effect of FIP and SOI timing on CRDI engine performance, emissions and combustion. • Fuel injection duration shortened, peak injection rate increased with increasing FIP. • SMD (D 32 ) and AMD (D 10 ) of fuel droplets decreased for lower biodiesel blends. • Increase in biodiesel blend ratio and FIP, fuel injection duration decreased. - Abstract: In this investigation, effect of 10%, 20% and 50% Karanja biodiesel blends on injection rate, atomization, engine performance, emissions and combustion characteristics of common rail direct injection (CRDI) type fuel injection system were evaluated in a single cylinder research engine at 300, 500, 750 and 1000 bar fuel injection pressures at different start of injection timings and constant engine speed of 1500 rpm. The duration of fuel injection slightly decreased with increasing blend ratio of biodiesel (Karanja Oil Methyl Ester: KOME) and significantly decreased with increasing fuel injection pressure. The injection rate profile and Sauter mean diameter (D 32 ) of the fuel droplets are influenced by the injection pressure. Increasing fuel injection pressure generally improves the thermal efficiency of the test fuels. Sauter mean diameter (D 32 ) and arithmetic mean diameter (D 10 ) decreased with decreasing Karanja biodiesel content in the blend and significantly increased for higher blends due to relatively higher fuel density and viscosity. Maximum thermal efficiency was observed at the same injection timing for biodiesel blends and mineral diesel. Lower Karanja biodiesel blends (up to 20%) showed lower brake specific hydrocarbon (BSHC) and carbon monoxide (BSCO) emissions in comparison to mineral diesel. For lower Karanja biodiesel blends, combustion duration was shorter than mineral diesel however at higher fuel injection pressures, combustion duration of 50% blend was longer than mineral diesel. Up to 10% Karanja biodiesel blends in a CRDI

Consideration of hot channel factors in design for providing operating margins on coolant channel outlet temperature

International Nuclear Information System (INIS)

Sharma, V.K.; Surendar, C.; Bapat, C.N.

1994-01-01

The Indian Pressurized Heavy Water Reactors (IPHWR) are horizontal pressure tube reactors using natural uranium oxide fuel in the form of short (495 mm) clusters. The fuel clusters in the Zr-Nb pressure tubes are cooled by high pressure, high temperature and subcooled circulating heavy water. Coolant flow distribution to individual channels is designed to match the power distribution so as to obtain uniform coolant outlet temperature. However, during operation, the coolant outlet temperature in individual channels deviate from their nominal value due to: tolerances in process design; effects of grid frequency on the pump speed; deviation in channel powers from the nominal values due to on-power fuelling and movement of reactivity devices, and so on. Thus an operating margin, between the highest permissible and nominal coolant outlet temperatures, is required taking into account various hot channel factors that contribute to higher coolant outlet temperatures. The paper discusses the methodology adopted to assess various hot channel factors which would provide optimum operating margins while ensuring sub-cooling. (author)

Low Blood Pressure

Science.gov (United States)

... a problem. Sometimes blood pressure that is too low can also cause problems. Blood pressure is the ... reading is 90/60 or lower, you have low blood pressure. Some people have low blood pressure ...

Continuous positive airway pressure breathing increases the spread of sensory blockade after low-thoracic epidural injection of lidocaine.

NARCIS (Netherlands)

Visser, W.A.; Gielen, M.J.M.; Giele, J.L.P.

2006-01-01

Factors affecting the distribution of sensory blockade after epidural injection of local anesthetics remain incompletely clarified. To evaluate if increasing intrathoracic pressure affects the spread of thoracic epidural anesthesia, we randomized 20 patients who received an epidural catheter at the

Study on the quench behavior of molten fuel material jet into coolant

International Nuclear Information System (INIS)

Abe, Yutaka; Kizu, Tetsuya; Arai, Takahiro; Nariai, Hideki; Chitose, Keiko; Koyama, Kazuya

2004-01-01

In a core disruptive accident (CDA) of a Fast Breeder Reactor, the post accident heat removal (PAHR) is crucial for the accident mitigation. The molten core material should be solidified in the sodium coolant in the reactor vessel. In the present experiment, molten material jet is injected into water to experimentally obtain fragments and the visualized information of the fragmentation. The distributed particle behavior of the molten material jet is observed with high-speed video camera. The distributions of the fragmented droplet diameter from the molten material jet are evaluated by correcting the solidified particles. The experimental results of the mean fragmented droplet diameter are compared with the existing theories. Consequently, the fragmented droplet diameter is close to the value estimated based on the Kelvin-Helmholtz instability. Once the particle diameter of the fragmented molten material could be known from a hydrodynamic model, it becomes possible to estimate the mass ratio of the molten particle to the total injected mass by combining an appropriate heat transfer model. The heat transfer model used in the present study is composed of the fragmentation model based on the Kelvin-Helmholtz instability. The mass ratio of the molten fragment to total mass of the melted mixed oxide fuel in sodium coolant estimated in the present study is very small. The result means that most of the molten mixed oxide fuel material injected into the sodium coolant can be cooled down under the solidified temperature, that is so called quenched, if the amount of the coolant is sufficient. (author)

Fuel-Coolant Interactions - some Basic Studies at the UKAEA Culham Laboratory

International Nuclear Information System (INIS)

Reynolds, J.A.; Dullforce, T.A.; Peckover, R.S.; Vaughan, G.J.

1976-01-01

In a hypothetical fault sequence important effects of fuel-coolant interactions include voiding and dispersion of core debris as well as the pressure damage usually discussed. The development of the fuel-coolant interaction probably depends on any pre-mixing Weber break-up that may occur, and is therefore a function of the way the fuel and coolant come together. Four contact modes are identified: jetting, shock tube, drops and static, and Culham's experiments have been mainly concerned with simulating the falling drop mode by using molten tin in water. It was observed that the fuel-coolant interaction is a short series of violent coolant oscillations centred at a localized position on the drop, generating a spray of submillimeter sized debris. The interaction started spontaneously at a specific time after the drop first contacted the water. There was a definite limited fuel-coolant interaction zone on a plot of initial coolant temperature versus initial fuel temperature outside which interactions never occurred. The. interaction time was a function of the initial temperatures. Theoretical scaling formulae are given which describe the fuel-coolant interaction zone and dwell time. Bounds of fuel and coolant temperature below which fuel-coolant interactions do not occur are explained by freezing. Upper bounds of fuel and coolant temperatures above which there were no fuel-coolant interactions are interpreted in terms of heat transfer through vapour films of various thicknesses. In conclusion: We have considered the effects of fuel-coolant interactions in a hypothetical fault sequence, emphasising that debris and vapour production as well as the pressure pulse can be important factors. The fuel-coolant interaction has been classified into types, according to possible modes of mixing in the fault sequence. Culham has been studying one type, the self-triggering of falling drops, by simulant experiments. It is found that there is a definite zone of interaction on a plot

Temporal pore pressure induced stress changes during injection and depletion

Science.gov (United States)

Müller, Birgit; Heidbach, Oliver; Schilling, Frank; Fuchs, Karl; Röckel, Thomas

2016-04-01

Induced seismicity is observed during injection of fluids in oil, gas or geothermal wells as a rather immediate response close to the injection wells due to the often high-rate pressurization. It was recognized even earlier in connection with more moderate rate injection of fluid waste on a longer time frame but higher induced event magnitudes. Today, injection-related induced seismicity significantly increased the number of events with M>3 in the Mid U.S. However, induced seismicity is also observed during production of fluids and gas, even years after the onset of production. E.g. in the Groningen gas field production was required to be reduced due to the increase in felt and damaging seismicity after more than 50 years of exploitation of that field. Thus, injection and production induced seismicity can cause severe impact in terms of hazard but also on economic measures. In order to understand the different onset times of induced seismicity we built a generic model to quantify the role of poro-elasticity processes with special emphasis on the factors time, regional crustal stress conditions and fault parameters for three case studies (injection into a low permeable crystalline rock, hydrothermal circulation and production of fluids). With this approach we consider the spatial and temporal variation of reservoir stress paths, the "early" injection-related induced events during stimulation and the "late" production induced ones. Furthermore, in dependence of the undisturbed in situ stress field conditions the stress tensor can change significantly due to injection and long-term production with changes of the tectonic stress regime in which previously not critically stressed faults could turn to be optimally oriented for fault reactivation.

Investigating skin penetration depth and shape following needle-free injection at different pressures: A cadaveric study.

Science.gov (United States)

Seok, Joon; Oh, Chang Taek; Kwon, Hyun Jung; Kwon, Tae Rin; Choi, Eun Ja; Choi, Sun Young; Mun, Seog Kyun; Han, Seung-Ho; Kim, Beom Joon; Kim, Myeung Nam

2016-08-01

amount of pressure used. When injecting 5% MB at 6 bars, the depth of infiltration reached 2.323 mm, while that at 8.5 bars reached 8.906 mm. The area of the hole created by the 5% MB injection was 0.797 mm(2) at 6 bars and 0.242 mm(2) at 8.5 bars. Latex injections reached a depth of 3.480 mm at 6 bars and 7.558 mm at 8.5 bars, and the areas were measured at 1.043 mm(2) (6 bars) and 0.355 mm(2) (8.5 bars). Histological examination showed that the injection penetrated as deep as the superficial musculoaponeurotic system at 6 bars and the masseter muscle at 8.5 bars. When injecting material into the skin using a pneumatic needle-free injector, higher-pressure injections result in a hole with smaller area than lower-pressure injections. The depth and shape of skin penetration vary according to the amount of pressure applied. For materials of low density and viscosity, there is a greater difference in penetration depth according to the degree of pressure. Lasers Surg. Med. 48:624-628, 2016. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

Reactor coolant pump for a nuclear reactor

International Nuclear Information System (INIS)

Burkhardt, W.; Richter, G.

1976-01-01

An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

Chemical aspects of hydrogen ingress in zirconium and zircaloy pressure tubes: ageing management of Indian PHWR coolant channels - determination of hydrogen and deuterium

International Nuclear Information System (INIS)

Sayi, Y.S.; Shankaran, P.S.; Yadav, C.S.; Ramanjaneyulu, P.S.; Venugopal, V.; Ramakumar, K.L.; Chhapru, G.C.; Prasad, R.; Jain, H.C.; Sood, D.D.

2009-02-01

Pressurized heavy water reactors (PHWRs) use zirconium and zirconium based alloys as clad and coolant tubes since its beginning. The first ever zircaloy-2 pressure tube failure occurred in 1983 at Ontario Hydro's Pickering Unit 2 in Canada which necessitated a thorough examination of causes of such failure. The failure was attributed to massive hydriding at the failed spot of pressure tube. Continuous usage of zirconium alloys could result in their hydrogen and deuterium pick-up leading to hydrogen/ deuterium embrittlement. The life of the zircaloy coolant channels is dictated by hydrogen/deuterium content and hence ageing management of the pressure tubes is essential for ensuring their trouble-free usage. It is desirable to have a sound knowledge on the chemical aspects of zirconium and zirconium based alloys metallurgy, the mechanistic principles of hydrogen ingress into the pressure tubes during in reactor service, and identifying suitable analytical methodologies for precise and accurate determination of hydrogen in wafer thin sliver samples carved out from insides of pressure tubes without causing any structural damage so that it can continue to remain in service. This is desirable so that the ageing management does not result in cost-escalation. This report is divided in to three main parts. The first part deals with the chemical aspects of zirconium and zirconium based alloy metallurgy, the mechanism of hydrogen pick-up and hydride formation in zirconium matrix. The second part describes various methodologies and their limitations, available for hydrogen/deuterium determination. The third part deals in detail, about the extensive investigations carried out at Radioanalytical Chemistry Division (RACD) in Radiochemistry and Isotope Group for establishing an indigenously developed hot vacuum extraction system in combination with quadrupole mass spectrometry for precise determination of hydrogen and deuterium in wafer thin sliver sample of zircaloy. The

Analysis of the behaviour of pressure and temperature of the containment of a PWR reactor, submitted to a postulated loss-of-coolant accident

International Nuclear Information System (INIS)

Silva, D.E. da; Arrieta, L.A.J.; Costa, J.R.; Camargo, C.; Santos, C.M. dos; Rochedo, E.R.R.

1979-12-01

The main purpose of this work is to analyse the pressure and temperature behaviour of the metalic containment of a PWR building, submitted to a postulated loss-of-coolant accident (LOCA) caused by a double-ended rupture in the main line of the primary circuit. The scope of the study was directed to verify the Final Safety Analysis Report (FSAR) results for the integrity of the metalic containment of the Angra I power plant. The highest containment pressure peak for this unit is expected for a break in the suction line of one of the main pumps of the primary coolant. Using the same input data, our results are very similar to those presented in the FSAR which shows a reasonable equivalence between the two analytical models. Using as input data the results of a previous LOCA study at CNEN, which yields to more conservative boundary conditions than those presented by the FSAR, the pressure and temperature peak values determined by our model are quite larger than those presented by the cited Safety Report. (author) [pt

Experimental verification of integrated pressure suppression systems in fusion reactors at in-vessel loss-of-coolant events

International Nuclear Information System (INIS)

Takase, K.; Akimoto, H.

2001-01-01

An integrated ICE (Ingress-of-Coolant Event) test facility was constructed to demonstrate that the ITER safety design approach and design parameters for the ICE events are adequate. Major objectives of the integrated ICE test facility are: to estimate the performance of an integrated pressure suppression system; to obtain the validation data for safety analysis codes; and to clarify the effects of two-phase pressure drop at a divertor and the direct-contact condensation in a suppression tank. A scaling factor between the test facility and ITER-FEAT is around 1/1600. The integrated ICE test facility simulates the ITER pressure suppression system and mainly consists of a plasma chamber, vacuum vessel, simulated divertor, relief pipe and suppression tank. From the experimental results it was found quantitatively that the ITER pressure suppression system is very effective to reduce the pressurization due to the ICE event. Furthermore, it was confirmed that the analytical results of the TRAC-PF1 code can simulate the experimental results with high accuracy. (author)

Experiment data report for LOFT nonnuclear test L1-3

International Nuclear Information System (INIS)

Millar, G.M.

1977-04-01

Test L1-3 was the third in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200 percent double-ended shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were: temperature at 540 0 F, pressure at 2256 psig, and loop flow at 2.34 x 10 6 lbm/hr. During system depressurization, emergency core cooling water was specified to be injected into the lower plenum of the reactor vessel using an accumulator, a low-pressure injection system pump, and a high-pressure injection system pump to provide data on the effects of emergency core cooling on the system thermal-hydraulic response. Injection into the lower plenum was initiated from the high- and low-pressure injection systems. Injection from the accumulator, however, was not initiated because a valve was inadvertently left closed. The experiment, therefore, was not completely successful in that one of the objectives outlined in the experiment operating specification for this test was not accomplished. Test L1-3 was repeated at Test L1-3A to meet the experimental requirements. Despite these difficulties, Test L1-3 did provide very valuable data to verify experiment repeatability

Physical properties of organic coolants

International Nuclear Information System (INIS)

Debbage, A.G.; Garton, D.A.; Kinneir, J.H.

1963-03-01

Density, viscosity, specific heat, vapour pressure and calorific value were measured within the temperature range 100 - 400 deg C for mixtures of Santowax R with pyrolytic high boiler and Santowax R with O.M.R.E. radiolytic high boiler; in addition measurements were made on Santowax OM, X-7 standard, X-7 loop coolant and O.M.R.E. coolant supplied by Atomic Energy of Canada Ltd. The accuracy of the measurements made were density (± 1/4%), viscosity (± 2%), specific heat (± 2%), vapour pressure (± 2%) and calorific value (± 1/2%). Thermal conductivity was calculated from an improved form of the Smiths equation with an accuracy within ± 6%. Equations fitted to the vapour pressure results were used to provide data outside the experimental range for burnout correlation purposes. The general effect of high boiler content on the specific heat and calorific values was small. The differences in physical property values for corresponding values of either pyrolytic or radiolytic high boiler were small for density (0.3%) and specific heat (2%), but quite large for viscosity (70%) with the pyrolytic high boiler mixture giving the higher value. The chemical analysis of all materials was based on gas chromatography and the relationship between this and an earlier distillation method established. (author)

Stress Analysis of Fuel Rod under Axial Coolant Flow

Energy Technology Data Exchange (ETDEWEB)

Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

2010-05-15

A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

Stress Analysis of Fuel Rod under Axial Coolant Flow

International Nuclear Information System (INIS)

Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

2010-01-01

A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

Alternate method of decay-heat removal in a C-E plant following a SBLOCA

International Nuclear Information System (INIS)

Boyack, B.E.

1986-01-01

The use of an atmospheric steam-dump procedure to cool and depressurize a Combustion-Engineering plant, Calvert Cliffs-1, following small-break loss-of-coolant accidents with failure of the high-pressure injection system to operate has been investigated. The procedure was effective in depressuizing the primary to the low-pressure injection system operating pressure and design temperature using water supplies from the safety-grade condensate water storage tank only. The procedure was found to be effective even if additional failures occurred. Specifically, low-pressure injection conditions were attained if only a single atmospheric dump valve was available or if the safety-injection tanks (accumulators) were not available

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.

Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no load conditions

Energy Technology Data Exchange (ETDEWEB)

Shao, Ge; Cao, Xuewu [School of Mechanical and Engineering, Shanghai Jiao Tong University, Shanghai (China)

2015-06-15

The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

Pressure and pressure derivative analysis for injection tests with variable temperature without type-curve matching

International Nuclear Information System (INIS)

Escobar, Freddy Humberto; Martinez, Javier Andres; Montealegre Matilde

2008-01-01

The analysis of injection tests under nonisothermic conditions is important for the accurate estimation of the reservoir permeability and the well's skin factor; since previously an isothermical system was assumed without taking into account a moving temperature front which expands with time plus the consequent changes in both viscosity and mobility between the cold and the hot zone of the reservoir which leads to unreliable estimation of the reservoir and well parameters. To construct the solution an analytical approach presented by Boughrara and Peres (2007) was used. That solution was initially introduced for the calculation of the injection pressure in an isothermic system. It was later modified by Boughrara and Reynolds (2007) to consider a system with variable temperature in vertical wells. In this work, the pressure response was obtained by numerical solution of the anisothermical model using the Gauss Quadrature method to solve the integrals, and assuming that both injection and reservoir temperatures were kept constant during the injection process and the water saturation is uniform throughout the reservoir. For interpretation purposes, a technique based upon the unique features of the pressure and pressure derivative curves were used without employing type-curve matching (TDS technique). The formulation was verified by its application to field and synthetic examples. As expected, increasing reservoir temperature causes a decrement in the mobility ratio, then estimation of reservoir permeability is some less accurate from the second radial flow, especially, as the mobility ratio increases

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.

Design of Safety Injection Tanks Using Axiomatic Design and TRIZ

Energy Technology Data Exchange (ETDEWEB)

Heo, Gyunyoung [Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do, 446-701 (Korea, Republic of); Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

2008-07-01

Design can be categorized into two steps: 'synthesis' and 'analysis'. While synthesis is the process of decision-making on design parameters, analysis is the process of optimizing the parameters selected. It is known from experience that the mistakes made in the synthesis process are hardly corrected in the analysis process. 'Systematic synthesis' is, therefore, easy to overlook but an important topic. 'Systematic' is interpreted as 'minimizing' uncertainty and subjectivity. This paper will introduce the design product achieved by using Axiomatic Design (AD) and TRIZ (Theory of Inventive Problem Solving romanized acronym for Russian), which is a new design of Safety Injection Tank (SIT). In designing a large-capacity SIT which should play an important role in mitigating the large break loss of coolant accidents, there are three issues: 1) the excessively large plenum for pressurized nitrogen gas; 2) the difficulties maintaining the high initial injection flow rate; and 3) the non-condensable nitrogen gas in the coolant. This study proposes a conceptual idea for SITs that are pressurized by the chemical reaction of solid propellants. The AD theory and the principles of TRIZ enable new approach in problem-solving for those three issues in an innovative way. The paper made an effort to clarify the systematic synthesis process to reach the final design solution. (authors)

Accident tolerant high-pressure helium injection system concept for light water reactors

International Nuclear Information System (INIS)

Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

2016-01-01

Highlights: • Potential helium injection strategy is proposed for LWR accident scenarios. • Multiple injection sites are proposed for current LWR designs. • Proof-of-concept experimentation illustrates potential helium injection benefits. • Computational studies show an increase in pressure vessel blowdown time. • Current LOCA codes have the capability to include helium for feasibility calculations. - Abstract: While the design of advanced accident-tolerant fuels and structural materials continues to remain the primary focus of much research and development pertaining to the integrity of nuclear systems, there is a need for a more immediate, simple, and practical improvement in the severe accident response of current emergency core cooling systems. Current blowdown and reflood methodologies under accident conditions still allow peak cladding temperatures to approach design limits and detrimentally affect the integrity of core components. A high-pressure helium injection concept is presented to enhance accident tolerance by increasing operator response time while maintaining lower peak cladding temperatures under design basis and beyond design basis scenarios. Multiple injection sites are proposed that can be adapted to current light water reactor designs to minimize the need for new infrastructure, and concept feasibility has been investigated through a combination of proof-of-concept experimentation and computational modeling. Proof-of-concept experiments show promising cooling potential using a high-pressure helium injection concept, while the developed choked-flow model shows core depressurization changes with added helium injection. Though the high-pressure helium injection concept shows promise, future research into the evaluation of system feasibility and economics are needed.Classification: L. Safety and risk analysis

Trace organics in AGR coolants

International Nuclear Information System (INIS)

Smith, R.; Green, L.O.; Johnson, P.A.V.

1980-01-01

Several analytical techniques have been employed in previous studies of the stable organic compounds arising from the radiolysis of methane/carbon monoxide/carbon dioxide coolants. The majority of this early information was collected from the Windscale AGR prototype. Analyses were also carried out on the liquors obtained from the WAGR humidryers. Three classes of compound were found in the liquors; aliphatic acids in the aqueous phase and methyl ketones and aromatic hydrocarbons in the oily phase. Acetic acid was found to be the predominant carboxylic acid. This paper outlines the major findings from a recent analytical survey of coolants taken over a wide range of dose rate, pressure, temperature and composition, from materials testing reactor facilities, WAGR and CAGR. (author)

Analysis of the loss of coolant accident due to the faiture in the open position of two pressurizer relief valves, for Angra-1 nuclear power plant

International Nuclear Information System (INIS)

Freire, C.F.

1981-06-01

A study of the modeling techniques adequate for simulating the loss of coolant accident caused by stuck open pressurizer relief valves, using the RELAP4-MOD5 code, is performed and the model developed is applied to the analysis of this kind of accident for the Central Nuclear Almirante Alvaro Alberto Unit (Angra 1). The thermal hydraulic behavior of the reactor cooling system, when subjected to a loss of main feedwater followed by the failure in the open position of two pressurizer relief valves, is determined. The relief valves are assumed to fail in the totally open position, delivering the maximum massflow through the discharge line. The RELAP4-MOD5 code is shown to be adequate for this kind of analysis, and the detailed prediction of the thermal hydraulic behavior of the Reactor Coolant System is thus possible. The eficiency of the emergency core cooling system of Angra 1 is demonstrated, the fuel elements remaining covered by the coolant during all the accident, and the peak clad temperatures are kept within design limites, ensuring the integrity of the core. (Author) [pt

An emergency water injection system (EWIS) for future CANDU reactors

International Nuclear Information System (INIS)

Marques, Andre L.F.; Todreas, Neil E.; Driscoll, Michael J.

2000-01-01

This paper deals with the investigation of the feasibility and effectiveness of water injection into the annulus between the calandria tubes and the pressure tubes of CANDU reactors. The purpose is to provide an efficient decay heat removal process that avoids permanent deformation of pressure tubes severe accident conditions, such as loss of coolant accident (LOCA). The water injection may present the benefit of cost reduction and better actuation of other related safety systems. The experimental work was conducted at the Massachusetts Institute of Technology (MIT), in a setup that simulated, as close as possible, a CANDU bundle annular configuration, with heat fluxes on the order of 90 kW/m 2 : the inner cylinder simulates the pressure tube and the outer tube represents the calandria tube. The experimental matrix had three dimensions: power level, annulus water level and boundary conditions. The results achieved overall heat transfer coefficients (U), which are comparable to those required (for nominal accident progression) to avoid pressure tube permanent deformation, considering current CANDU reactor data. Nonetheless, future work should be carried out to investigate the fluid dynamics such as blowdown behavior, in the peak bundle, and the system lay-out inside the containment to provide fast water injection. (author)

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)

Primary coolant recycling device for FBR type reactor

International Nuclear Information System (INIS)

Kanbe, Mitsuru; Tokiwai, Moriyasu

1998-01-01

A primary coolants (liquid sodium) recycling device comprises a plurality of recycling pumps. The recycling pumps are operated while using, as a power source, electric power generated by a thermoelectric power generation system by utilizing heat stored in the coolants. The thermoelectric power generation system comprises a thermo-electric conversion module, heat collecting heat pipes as a high temperature side heat conduction means and heat dissipating pipes as a low temperature side heat conduction means. The heat of coolants is transferred to the surface of the high temperature side of each thermo-electric conversion elements of the thermal power generation system by the heat collecting heat pipes. The heat on the low temperature side of each of the thermo-electric conversion elements is removed by the heat dissipating pipes. Accordingly, temperature difference is caused between both surfaces of the thermo-electric conversion elements. Even upon loss of a main power source due to stoppage of electricity, electric power is generated by utilizing heat of coolants, so that the recycling pumps circulate coolants to cool a reactor core continuously. (I.N.)

Investigating Liquid CO2 as a Coolant for a MTSA Heat Exchanger Design

Science.gov (United States)

Paul, Heather L.; Padilla, Sebastian; Powers, Aaron; Iacomini, Christie

2009-01-01

Metabolic heat regenerated Temperature Swing Adsorption (MTSA) technology is being developed for thermal and carbon dioxide (CO 2) control for a future Portable Life Support System (PLSS), as well as water recycling. CO 2 removal and rejection is accomplished by driving a sorbent through a temperature swing of approximately 210 K to 280 K . The sorbent is cooled to these sub-freezing temperatures by a Sublimating Heat Exchanger (SHX) with liquid coolant expanded to sublimation temperatures. Water is the baseline coolant available on the moon, and if used, provides a competitive solution to the current baseline PLSS schematic. Liquid CO2 (LCO2) is another non-cryogenic coolant readily available from Martian resources which can be produced and stored using relatively low power and minimal infrastructure. LCO 2 expands from high pressure liquid (5800 kPa) to Mars ambient (0.8 kPa) to produce a gas / solid mixture at temperatures as low as 156 K. Analysis and experimental work are presented to investigate factors that drive the design of a heat exchanger to effectively use this sink. Emphasis is given to enabling efficient use of the CO 2 cooling potential and mitigation of heat exchanger clogging due to solid formation. Minimizing mass and size as well as coolant delivery are also considered. The analysis and experimental work is specifically performed in an MTSA-like application to enable higher fidelity modeling for future optimization of a SHX design. In doing so, the work also demonstrates principles and concepts so that the design can be further optimized later in integrated applications (including Lunar application where water might be a choice of coolant).

Concordant pressure paresthesia during interlaminar lumbar epidural steroid injections correlates with pain relief in patients with unilateral radicular pain.

Science.gov (United States)

Candido, Kenneth D; Rana, Maunak V; Sauer, Ruben; Chupatanakul, Lalida; Tharian, Antony; Vasic, Vladimir; Knezevic, Nebojsa Nick

2013-01-01

Transforaminal and interlaminar epidural steroid injections are commonly used interventional pain management procedures in the treatment of radicular low back pain. Even though several studies have shown that transforaminal injections provide enhanced short-term outcomes in patients with radicular and low back pain, they have also been associated with a higher incidence of unintentional intravascular injection and often dire consequences than have interlaminar injections. We compared 2 different approaches, midline and lateral parasagittal, of lumbar interlaminar epidural steroid injection (LESI) in patients with unilateral lumbosacral radiculopathic pain. We also tested the role of concordant pressure paresthesia occurring during LESI as a prognostic factor in determining the efficacy of LESI. Prospective, randomized, blinded study. Pain management center, part of a teaching-community hospital in a major metropolitan US city. After Institutional Review Board approval, 106 patients undergoing LESI for radicular low back pain were randomly assigned to one of 2 groups (53 patients each) based on approach: midline interlaminar (MIL) and lateral parasagittal interlaminar (PIL). Patients were asked to grade any pressure paresthesia as occurring ipsilaterally or contralaterally to their "usual and customary pain," or in a distribution atypical of their daily pain. Other variables such as: the Oswestry Disability Index questionnaire, pain scores at rest and during movement, use of pain medications, etc. were recorded 20 minutes before the procedure, and on days 1, 7, 14, 21, 28, 60, 120, 180 and 365 after the injection. Results of this study showed statistically and clinically significant pain relief in patients undergoing LESI by both the MIL and PIL approaches. Patients receiving LESI using the lateral parasagittal approach had statistically and clinically longer pain relief then patients receiving LESI via a midline approach. They also had slightly better quality of

Pressure changes in the plasma sheet during substorm injections

International Nuclear Information System (INIS)

Kistler, L.M.; Moebuis, E.; Baumjohann, W.; Paschmann, G.; Hamilton, D.C.

1992-01-01

The authors have determined the particle pressure and total pressure as a function of radial distance in the plasma sheet for periods before and after the onset of substorm-associated ion enhancements over the radial range 7-19 R E . They have chosen events occurring during times of increasing magnetospheric activity, as determined by an increasing AE index, in which a sudden increase, or injection, of energetic particle flux is observed. During these events the particle energy of maximum contribution to the pressure increases from about 12 to about 27 keV. In addition, the particle pressure increases, and the magnetic pressure decreases, with the total pressure only changing slightly. For radial distances of less than 10 R E the total pressure tends to increase with the injection, while outside 10 R E it tends to decrease or remain the same. Because the fraction of the pressure due to particles has increased and higher energies are contributing to the pressure, a radial gradient is evident in the postinjection, but not preinjection, flux measurements. These observations show that the simulations appearance of energetic particles and changes in the magnetic field results naturally from pressure balance and does not necessarily indicate that the local changing field is accelerating the particles. The changes in the total pressure outside 10 R E are consistent with previous measurements of pressure changes at substorm onset and can be understood in terms of the unloading of energy in the magnetotail and the resulting change in the magnetic field configuration

Efficiency of water coolant for DEMO divertor

International Nuclear Information System (INIS)

Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

2015-01-01

Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

Efficiency of water coolant for DEMO divertor

Energy Technology Data Exchange (ETDEWEB)

Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

2015-10-15

Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

Heat transfer and fluid flow aspects of fuel--coolant interactions

International Nuclear Information System (INIS)

Corradini, M.L.

1978-09-01

A major portion of the safety analysis effort for the LMFBR is involved in assessing the consequences of a Hypothetical Core Disruptive Accident (HCDA). The thermal interaction of the hot fuel and the sodium coolant during the HCDA is investigated in two areas. A postulated loss of flow transient may produce a two-phase fuel at high pressures. The thermal interaction phenomena between fuel and coolant as the fuel is ejected into the upper plenum are investigated. A postulated transient overpower accident may produce molten fuel being released into sodium coolant in the core region. An energetic coolant vapor explosion for these reactor materials does not seem likely. However, experiments using other materials (e.g., Freon/water, tin/water) have demonstrated the possibility of this phenomenon

Method of determination of thermo-acoustic coolant instability boundaries in reactor core at NPPs with WWER

International Nuclear Information System (INIS)

Skalozubov, Volodymyr; Kolykhanov, Viktor; Kovryzhkin, Yuriy

2007-01-01

The regulatory body of Ukraine, the National Atomic Energy Company and the Scientific and Production Centre have led team-works concerned with previously unstudied factors or phenomena affecting reactor safety. As a result it is determined that the thermo-acoustic coolant instability conditions can appear in the core at definite operating WWER regimes. Considerable cyclic dynamic loads affect fuel claddings over thermo-acoustic pressure oscillations. These loads can result in inadmissible cassette design damage and containment damage. Taking into account calculation and experimental research authors submit a method of on-line assessment of WWER core state concerning thermo-acoustic coolant instability. According to this method, the thermo-acoustic coolant instability appearance conditions can be estimated using normal registered parameters (pressure, temperature, heat demand etc.). At operative modes, a WWER-1000 core is stable to tracheotomies oscillations, but reduction of coolant discharge through the core for some times can result in thermo-acoustic coolant instability. Thermo-acoustic instability appears at separate transitional modes concerned with reactor scram and unloading/loading at all power units. When thermo-acoustic instability begins in transitional modes, core elements are under influence of high-frequency coolant pressure pulsations for a long time (tens of hours)

Influence of injection pressures till to 1,000 bar on the carburetion in a spark ignition engine with direct injection; Einfluss von Einspritzdruecken bis 1000 bar auf die Gemischbildung in einem Ottomotor mit Direkteinspritzung

Energy Technology Data Exchange (ETDEWEB)

Buri, Stefan; Schumann, Florian; Kubach, Heiko; Spicher, Ulrich [Karlsruher Institut fuer Technologie (KIT) (DE). Inst. fuer Kolbenmaschinen (IFKM); Kneifel, Alexander [MTU Friedrichshafen GmbH (Germany)

2011-07-01

This paper presents the results of optical investigations of the impact of injection pressures of up to 1000 bar on mixture formation in a spray-guided direct injection engine. The maximum load in stratified operation of an engine with such a spray-guided combustion system is limited by the achievable quality of the mixture. In particular, when using multi hole injectors, the limit of stratified operation is reached rather early, due to comparatively low flow rates and thus insufficient stratification. One measure to increase the flow rate is to increase the injection pressure. The goal of this measure is to generate a more compact stratification, leading to combustion at richer air fuel ratios. This enables reductions of burning duration, hydrocarbon- and particulate emissions. The fundamental impact of increasing the injection pressure from 200 up to 1000 bar on mixture formation was investigated by using LIF- and Mie-scattering in a pressure chamber. Following that, the mixture formation was investigated under real conditions in a single cylinder engine by visualizing the injection process using Mie-scattering. Finally the results of engine operation are compared with those from the pressure chamber. (orig.)

High-Pressure Injection Injuries to the Hand

Directory of Open Access Journals (Sweden)

Davod Jafari

2016-07-01

Full Text Available Background High-pressure injections into the hand, burden devastating and permanent functional impairments. Many materials including paint, paint thinner, gasoline, oil and grease are reported as the causative agents. These injuries need multiple procedures and reconstructions most of the time and 40% of the injuries may end with amputation of the injured part. Objectives The aim of this study was to report the treatment outcomes and methods of treatments of patients with high-pressure injection injuries of the hand. Methods We retrospectively reviewed the medical records, imaging files and demographic data of patients, who were treated at our center due to the high-pressure injuries to their hands. We recorded the kind of the injected materials, time to the first treatment procedure, times of operation, and methods of their treatments. The outcomes of the injuries as well as the deficiency of the digital joints motion were also reported. Results Nine cases with high-pressure injury of the hand were enrolled in this study. All patients were male with mean age of 26.88 ± 7.52. Mean follow-up time was 28.55 ± 12.49 months. The dominant hand was the right side in seven patients and left in two patients. Injury was in the left hand of seven patients and right hand of two patients. Index finger was the most common involved part (five cases followed by the thumb (two cases. Injected material was grease in seven cases, water-base paint and water, each in one case.Mean time delay to the first treatment procedure was 29.16 ± 25.66 hours for seven patients. This was exceptionally long for two patients (seven days and 24 months. Type of treatment was debridement and skin graft for three cases, debridement and cross finger flap for two cases, debridement for two cases and nerve graft for one case. Amputation of the necrotic digit was performed for one case. Mean hospitalization time was 8.33 ± 3.64 days for all patients.Mean total active range of motion

Rupture behaviour of nuclear fuel cladding during loss-of-coolant accident

Energy Technology Data Exchange (ETDEWEB)

Suman, Siddharth [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Khan, Mohd Kaleem, E-mail: mkkhan@iitp.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Pathak, Manabendra [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Singh, R.N.; Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

2016-10-15

Highlights: • Modelling of nuclear fuel cladding during loss-of-coolant accident transient. • Phase transformation, corrosion, and creep combined to evaluate burst criterion. • Effect of oxygen concentration on burst stress and burst strain. • Effect of heating rate, internal pressure fluctuation, shear modulus incorporated. - Abstract: A burst criterion model accounting the simultaneous phenomena of corrosion, solute-strengthening effect of oxygen, oxygen concentration based non-isothermal phase transformation, and thermal creep has been developed to predict the rupture behaviour of zircaloy-4 nuclear fuel cladding during the loss-of-coolant accident transients. The present burst criterion model has been validated using experimental data obtained from single-rod transient burst tests performed in steam environment. The predictions are in good agreement with the experimental results. A detailed computational analysis has been performed to assess the role of different parameters in the rupture of zircaloy cladding during loss-of-coolant accidents. This model reveals that at low temperatures, lower heating rates produce higher burst strains as oxidation effect is nominal. For high temperatures, the lower heating rates produce less burst strains, whereas higher heating rates yield greater burst strains.

Reactor Coolant Pump seal issues and their applicability to new reactor designs

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.

1993-01-01

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970's and early 1980's raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

Reactor coolant pump seal issues and their applicability to new reactor designs

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.

1993-01-01

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at U.S. operating plants during the 1970's and early 1980's raised concerns from the U.S. Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

The light water integral reactor with natural circulation of the coolant at supercritical pressure B-500 SKDI

International Nuclear Information System (INIS)

Silin, V.A.; Voznesensky, V.A.; Afrov, A.M.

1993-01-01

Pressure increase in the primary circuit over the critical value gives a possibility to construct the B-500SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in the vessel with a diameter less than 5 m. The given reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower as compared to a conventional PWR. The development of the reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology. (orig.)

Low Blood Pressure (Hypotension)

Science.gov (United States)

... lowest at night and rises sharply on waking. Blood pressure: How low can you go? What's considered low ... low blood pressure. Medications that can cause low blood pressure Some medications can cause low blood pressure, including: ...

REMIX: a computer program for temperature transients due to high pressure injection after interruption of natural circulation

International Nuclear Information System (INIS)

Iyer, K.; Nourbakhsh, H.P.; Theofanous, T.G.

1986-05-01

This report describes the features and use of several computer programs developed on the basis of the Regional Mixing Model (RMM). This model provides a phenomenologically-based analytical description of the stratified flow and temperature fields resulting from High Pressure Safety Injection (HPI) in the stagnated loops of a Pressurized Water Reactor (PWR). The basic program is called REMIX and is intended for thermally-induced stratification at low Froude number injections. The REMIX-S version is intended for solute-induced stratification with or without thermal effects as found in several experimental simulations. The NEWMIX program is a derivative of REMIX representing the limit of maximum possible mixing within the cold leg and is intended for high Froude number injections. The NEWMIX-S version accounts for solute effects. Listings of all programs and sample problem input and output files are included. 10 refs

Numerical investigation on critical heat flux and coolant volume required for transpiration cooling with phase change

International Nuclear Information System (INIS)

He, Fei; Wang, Jianhua

2014-01-01

Highlights: • Five states during the transpiration cooling are discussed. • A suit of applicable program is developed. • The variations of the thickness of two-phase region and the pressure are analyzed. • The relationship between heat flux and coolant mass flow rate is presented. • An approach is given to define the desired case of transpiration cooling. - Abstract: The mechanism of transpiration cooling with liquid phase change is numerically investigated to protect the thermal structure exposed to extremely high heat flux. According to the results of theoretical analysis, there is a lower critical and an upper critical external heat flux corresponding a certain coolant mass flow rate, between the two critical values, the phase change of liquid coolant occurs within porous structure. A strongly applicable self-edit program is developed to solve the states of fluid flow and heat transfer probably occurring during the phase change procedure. The distributions of temperature and saturation in these states are presented. The variations of the thickness of two-phase region and the pressure including capillary are analyzed, and capillary pressure is found to be the main factor causing pressure change. From the relationships between the external heat flux and coolant mass flow rate obtained at different cooling cases, an approach is given to estimate the maximal heat flux afforded and the minimal coolant consumption required by the desired case of transpiration cooling. Thus the pressure and coolant consumption required in a certain thermal circumstance can be determined, which are important in the practical application of transpiration cooling

High pressure injection injuries: an overview.

Science.gov (United States)

Fialkov, J A; Freiberg, A

1991-01-01

Injuries resulting from the use of high pressure injectors and spray guns are relatively rare; however, the potential tissue damage caused by the injury as well as the extent of the injury itself may go unrecognized by the primary physician. The purpose of this paper is to inform the emergency physician of the nature and standard management of this type of injury. A basic understanding of the pathophysiology of the high pressure injection injury (HPII) is essential in avoiding the mistakes in management that have been reported in the literature. The emergency management of the HPII includes: evaluation and immobilization, tetanus and antimicrobial prophylaxis, supportive and resuscitative measures, analgesia, and minimizing the time to definitive surgical treatment.

Analysis of the effects of the pressure wave generated in loss of coolant accidents in reactor vessels

International Nuclear Information System (INIS)

Valero Martinez, M.

1980-01-01

The increasing demands in the field of ''Nuclear Safety'', obliges to a perfect knowledge of the causes and effects of every possible accident in a nuclear power plant. In this paper will be analysed the effects of the pressure wave appearing in a LOCA (Loss of collant accident). The pressure wave could deform the following structures: core barrel wall, cover and bottom, control rods and safety coolant system. Any change of the geometry of these structures could provoke and incorrect system reaction after the accident has happened. The basis and hypothesis for the theoretical analysis will be exposed. The structures are considered to be rigid. A typical boiling water be analysed and the developed theory will be verified in comparations with experimental results and the results obtained with some others models. Due to the easy application and short calculation time of the created programmes, they are recommended for parametrical calculations in the analysis of the pressurized water reactors and boiling water reactors. (author)

Technical findings related to Generic Issue 23: Reactor coolant pump seal failure

International Nuclear Information System (INIS)

Ruger, C.J.; Luckas, W.J. Jr.

1989-03-01

Reactor coolant pumps contain mechanical seals to limit the leakage of pressurized coolant from the reactor coolant system to the containment. These seals have the potential to leak, and a few have degraded and even failed resulting in a small break loss of coolant accident (LOCA). As a result, ''Reactor Coolant Pump Seal Failure,'' Generic Issue 23 was established. This report summarizes the findings of a technical investigation generated as part of the program to resolve this issue. These technical findings address the various fact-finding issue tasks developed for the action plan associated with the generic issue, namely background information on seal failure, evaluation of seal cooling, and mechanical- and maintenance-induced failure mechanisms. 46 refs., 15 figs., 14 tabs

System and method for determining coolant level and flow velocity in a nuclear reactor

Science.gov (United States)

Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

2013-09-10

A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

A concept of JAERI passive safety light water reactor system (JPSR)

Energy Technology Data Exchange (ETDEWEB)

Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)

1995-09-01

The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

The dynamic pressure measurements of the nuclear reactor coolant for condition-based maintenance of the reactor

International Nuclear Information System (INIS)

Es-Saheb, M.H.H.

1990-01-01

The condition-based maintenance of the nuclear reactor, by monitoring and measuring the instantaneous dynamic pressure distribution of the coolant (water) impact on the solid surfaces of the reactor during operation is presented. The behaviour of water domes (jets) produced by underwater explosions of small changes of P.E.T.N. at various depths in two different size cylindrical containers, which simulate the nuclear reactor, is investigated. Water surface domes (jets) from the underwater explosions are photographed. Depending on the depth of the charge, curved and flat top jets of up to 455 mm diameter and impact speeds of up to 70 m/sec. are observed. The instabilities in the dome surfaces are observed and the instantaneous profiles are analysed. It is found that, in all cases tested, the maximum pressure takes place at the center of the jet and could reach up to 3.0 times the on-dimensional impact pressure value. The use of their measurements, as online monitoring for condition-based maintenance and design-out maintenance is discussed. 18 refs

Research and application of zinc injection in PWRs

International Nuclear Information System (INIS)

Jiang Lei

2012-01-01

In the middle 1990s, some PWRs in USA and Germany started to inject Zinc into the reactor coolant system for reducing both radiation fields and primary water stress corrosion cracking (PWSCC). Based on data from the labs and experience in the demonstration pants, Zinc injection obviously reduced radiation fields, and effectively mitigated PWSCC. Plants in USA injected high concentration zinc that is 15 ppb to 40 ppb to restrained PWSCC. Whereas, plants in Germany injected low concentration zinc that is 5 ppb to 10 ppb to reduce radiation fields. There are more than ten years at aspect of zinc rejection in overseas PWR, but domestic plants don't add zinc. The building PWR in Zhejiang Sanmen is the first AP1000 unit in the world, according to requirement of designers, it will start to inject zinc in the initial fuel cycle. (author)

Specificities of reactor coolant pumps units with lead and lead-bismuth coolant

International Nuclear Information System (INIS)

Beznosov, A.V.; Anotonenkov, M.A.; Bokov, P.A.; Baranova, V.S.; Kustov, M.S.

2009-01-01

The analysis results of impact of lead and lead-bismuth coolants specific properties on the coolants flow features in flow channels of the main and auxiliary circulating pumps are presented. Impossibility of cavitation initiation in flow channels of vane pumps pumping lead and lead-bismuth coolants was demonstrated. The experimental research results of discontinuity of heavy liquid metal coolant column were presented and conditions of gas cavitation initiation in coolant flow were discussed. Invalidity of traditional calculation methods of water and sodium coolants circulation pumps calculations for lead and lead-bismuth coolants circulation pumps was substantiated [ru

Comparison of MELCOR modeling techniques and effects of vessel water injection on a low-pressure, short-term, station blackout at the Grand Gulf Nuclear Station

International Nuclear Information System (INIS)

Carbajo, J.J.

1995-06-01

A fully qualified, best-estimate MELCOR deck has been prepared for the Grand Gulf Nuclear Station and has been run using MELCOR 1.8.3 (1.8 PN) for a low-pressure, short-term, station blackout severe accident. The same severe accident sequence has been run with the same MELCOR version for the same plant using the deck prepared during the NUREG-1150 study. A third run was also completed with the best-estimate deck but without the Lower Plenum Debris Bed (BH) Package to model the lower plenum. The results from the three runs have been compared, and substantial differences have been found. The timing of important events is shorter, and the calculated source terms are in most cases larger for the NUREG-1150 deck results. However, some of the source terms calculated by the NUREG-1150 deck are not conservative when compared to the best-estimate deck results. These results identified some deficiencies in the NUREG-1150 model of the Grand Gulf Nuclear Station. Injection recovery sequences have also been simulated by injecting water into the vessel after core relocation started. This marks the first use of the new BH Package of MELCOR to investigate the effects of water addition to a lower plenum debris bed. The calculated results indicate that vessel failure can be prevented by injecting water at a sufficiently early stage. No pressure spikes in the vessel were predicted during the water injection. The MELCOR code has proven to be a useful tool for severe accident management strategies

Direct injection of high pressure gas : scaling properties of pulsed turbulent jets

NARCIS (Netherlands)

Baert, R.S.G.; Klaassen, A.; Doosje, E.

2010-01-01

Existing gasoline DI injection equipment has been modified to generate single hole pulsed gas jets. Injection experiments have been performed at combinations of 3 different pressure ratios (2 of which supercritical) respectively 3 different hole geometries (i.e. length to diameter ratios). Injection

Method of charging instruments into liquid metal coolant

International Nuclear Information System (INIS)

Yamazaki, Hiroshi

1980-01-01

Purpose: To alleviate the thermal shock of a reactor charging machine when charging the machine into liquid metal coolant after the machine is preheated in cover gas. Method: When a reactor fueling machine reaches at the lowermost portion the position immediately above liquid metal coolant surface level, the machine is stopped moving down. The reactor fueling machine is heated at the lowermost portion by thermal radiation from the surface of the liquid metal coolant. After the machine is thus preheated in cover gas, it is again steadily moved down by a winch and charged into the liquid metal coolant. Therefore, the thermal shock of the machine becomes low when charging the machine into the liquid metal coolant to eliminate the damage and deformation at the machine. (Yoshihara, H.)

Effect of cavitation in high-pressure direct injection

Science.gov (United States)

Aboulhasanzadeh, Bahman; Johnsen, Eric

2015-11-01

As we move toward higher pressures for Gasoline Direct Injection and Diesel Direct Injection, cavitation has become an important issue. To better understand the effect of cavitation on the nozzle flow and primary atomization, we use a high-order accurate Discontinuous Galerkin approach using multi-GPU parallelism to simulate the compressible flow inside and outside the nozzle. Phase change is included using the six-equations model. We investigate the effect of nozzle geometry on cavitation inside the injector and on primary atomization outside the nozzle.

Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

Energy Technology Data Exchange (ETDEWEB)

Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

1995-09-01

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

Influence of contrast media viscosity and temperature on injection pressure in computed tomographic angiography: a phantom study.

Science.gov (United States)

Kok, Madeleine; Mihl, Casper; Mingels, Alma A; Kietselaer, Bas L; Mühlenbruch, Georg; Seehofnerova, Anna; Wildberger, Joachim E; Das, Marco

2014-04-01

Iodinated contrast media (CM) in computed tomographic angiography is characterized by its concentration and, consecutively, by its viscosity. Viscosity itself is directly influenced by temperature, which will furthermore affect injection pressure. Therefore, the purposes of this study were to systematically evaluate the viscosity of different CM at different temperatures and to assess their impact on injection pressure in a circulation phantom. Initially, viscosity of different contrast media concentrations (240, 300, 370, and 400 mgI/mL) was measured at different temperatures (20°C-40°C) with a commercially available viscosimeter. In the next step, a circulation phantom with physical conditions was used. Contrast media were prepared at different temperatures (20°C, 30°C, 37°C) and injected through a standard 18-gauge needle. All other relevant parameters were kept constant (iodine delivery rate, 1.9 g I/s; total amount of iodine, 15 g I). Peak flow rate (in milliliter per second) and injection pressure (psi) were monitored. Differences in significance were tested using the Kruskal-Wallis test (Statistical Package for the Social Sciences). Viscosities for iodinated CM of 240, 300, 370, and 400 mg I/mL at 20°C were 5.1, 9.1, 21.2, and 28.8 mPa.s, respectively, whereas, at 40°C, these were substantially lower (2.8, 4.4, 8.7, and 11.2 mPa.s). In the circulation phantom, mean (SD) peak pressures for CM of 240 mg I/mL at 20°C, 30°C, and 37°C were 107 (1.5), 95 (0.6), and 92 (2.1) psi; for CM of 300 mg I/mL, 119 (1.5), 104 (0.6), and 100 (3.6) psi; for CM of 370 mg I/mL, 150 (0.6), 133 (4.4), and 120 (3.5) psi; and for CM of 400 mg I/mL, 169 (1.0), 140 (2.1), and 135 (2.9) psi, respectively, with all P values less than 0.05. Low concentration, low viscosity, and high temperatures of CM are beneficial in terms of injection pressure. This should also be considered for individually tailored contrast protocols in daily routine scanning.

Comparative Experiments to Assess the Effects of Accumulator Nitrogen Injection on Passive Core Cooling During Small Break LOCA

Directory of Open Access Journals (Sweden)

Li Yuquan

2017-02-01

Full Text Available The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA, the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the

Comparative experiments to assess the effects of accumulator nitrogen injection on passive core cooling during small break LOCA

Energy Technology Data Exchange (ETDEWEB)

Li, YuQuan; Hao, Botao; Zhong, Jia; Wan Nam [State Nuclear Power Technology R and D Center, South Park, Beijing Future Science and Technology City, Beijing (China)

2017-02-15

The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME)—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative

Corrective actions to gas accumulation in safety injection system pipings of PWRs and gas void detection method

International Nuclear Information System (INIS)

Maki, Nobuo

2000-01-01

In the US, gas accumulation events of safety injection systems of PWRs during plant operation are continuously reported. As the events may result in loss of safety function, the USNRC is alerting licensees by Information Notices. The cause of the events is coolant leakage to interfacing systems with lower pressure, or gas dissolution of primary coolant by partial pressure drop. In this study, it was clarified by the evaluation of the cause of the events of US plants, gas accumulation in piping between an accumulator and Residual Heat Removal System should be quantitatively investigated regarding Japanese plants. Also, effectiveness of ultrasonic testing which is used for monthly gas accumulation surveillance in US plants was demonstrated using a model loop. In addition, the method was confirmed applicable by an experiment carried out at INSS to detect cavitation voids in piping systems. (author)

Evaluation of the gravity-injection capability for core cooling after a loss-of-SDC event

International Nuclear Information System (INIS)

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

1999-01-01

In order to evaluate the gravity-drain capability to maintain core cooling after a loss-of-shutdown-cooling event during shutdown operation, the plant conditions of the Young Gwang Units 3 and 4 were reviewed. The six cases of possible gravity-drain paths using the water of the refueling water storage tank (RWST) were identified and the thermal hydraulic analyses were performed using RELAP5/MOD3.2 code. The core cooling capability was dependent on the gravity-drain paths and the drain rate. In the cases with the injection path and opening on the different leg side, the system was well depressurized after gravity-injection and the core boiling was successfully prevented for a long-term transient. However, in the cases with the injection path and opening on the cold leg side, the core coolant continued boiling although the system pressure remains atmospheric after gravity-injection because the cold water injected from the RWST was bypassed the core region. In the cases with the higher pressurizer opening than the RWST water level, the system was also pressurized by the water-hold in the pressurizer and the core was uncovered because the gravity-injection from the RWST stopped due to the high system pressure. In addition, from the sensitivity study on the gravity-injection flow rates, it was found that about 54 kg/s of RWST drain rate was required to maintain the core cooling. Those analysis results would provide useful information to operators coping with the event

Application of the extended Kalman filtering for the estimation of core coolant flow rate in pressurized water reactors

International Nuclear Information System (INIS)

Shieh, D.J.; Upadhyaya, B.R.

1986-01-01

In-core neutron detector and core-exit temperature signals in a pressurized water reactor (PWR) satisfy the condition of observability of the core dynamic system, and can be used to estimate nonmeasurable state variables and model parameters. The extension of the Kalman filtering technique is very useful for direct parameter estimation. This approach is applied to the determination of core coolant mass flow rate in PWRs and is evaluated using in-core measurements at the Loss-of-Fluid Test (LOFT) reactor. The influence of model uncertainties on the estimation accuracy was studied using the ambiguity function analysis. A sequential discretization method was developed to achieve faster convergence to the true value, avoiding model discretization at each sample point. The performance of the extended Kalman filter and the computational innovations were evaluated using a reduced order core dynamic model of the LOFT reactor and random data simulation. The technique was then applied to the determination of LOFT core coolant flow rate from operational data at 100% and 65% flow conditions

Energy efficiency of a direct-injection internal combustion engine with high-pressure methanol steam reforming

International Nuclear Information System (INIS)

Poran, Arnon; Tartakovsky, Leonid

2015-01-01

This article discusses the concept of a direct-injection ICE (internal combustion engine) with thermo-chemical recuperation realized through SRM (steam reforming of methanol). It is shown that the energy required to compress the reformate gas prior to its injection into the cylinder is substantial and has to be accounted for. Results of the analysis prove that the method of reformate direct-injection is unviable when the reforming is carried-out under atmospheric pressure. To reduce the energy penalty resulted from the gas compression, it is suggested to implement a high-pressure reforming process. Effects of the injection timing and the injector's flow area on the ICE-SRM system's fuel conversion efficiency are studied. The significance of cooling the reforming products prior to their injection into the engine-cylinder is demonstrated. We show that a direct-injection ICE with high-pressure SRM is feasible and provides a potential for significant efficiency improvement. Development of injectors with greater flow area shall contribute to further efficiency improvements. - Highlights: • Energy needed to compress the reformate is substantial and has to be accounted for. • Reformate direct-injection is unviable if reforming is done at atmospheric pressure. • Direct-injection engine with high-pressure methanol reforming is feasible. • Efficiency improvement by 12–14% compared with a gasoline-fed engine was shown

Review on research of small break loss of coolant accident

International Nuclear Information System (INIS)

Bo Jinhai; Wang Fei

1998-01-01

The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

Linear titration plot for the determination of boron in the primary coolant of a pressurized water reactor

International Nuclear Information System (INIS)

Midgley, D.; Gatford, C.

1992-01-01

A linear titration plot method has been devised for the determination of boron as boric acid in partly neutralized solution, such as occurs in the primary coolant of pressurized water reactors. The total boron and the alkali in the sample are determined simultaneously. Although it is not essential to add mannitol in this method, it is more accurate when the solution is saturated with mannitol. Comparisons are made with other modes of titration: Gran plots, first and second differential potentiometric titrations and indicator titrations. None of these gives the total boron directly in partly neutralized solutions. (author)

Model Study of the Pressure Build-Up during Subcutaneous Injection

DEFF Research Database (Denmark)

Thomsen, Maria; Hernandez Garcia, Anier; Mathiesen, Joachim

2014-01-01

In this study we estimate the subcutaneous tissue counter pressure during drug infusion from a series of injections of insulin in type 2 diabetic patients using a non-invasive method. We construct a model for the pressure evolution in subcutaneous tissue based on mass continuity and the flow laws...

Breakdown pressures and characteristic flaw sizes during fluid injection experiments in shale at elevated confining pressures.

Science.gov (United States)

Chandler, M.; Mecklenburgh, J.; Rutter, E. H.; Taylor, R.; Fauchille, A. L.; Ma, L.; Lee, P. D.

2017-12-01

Fracture propagation trajectories in gas-bearing shales depend on the interaction between the anisotropic mechanical properties of the shale and the anisotropic in-situ stress field. However, there is a general paucity of available experimental data on their anisotropic mechanical, physical and fluid-flow properties, especially at elevated confining pressures. A suite of mechanical, flow and elastic measurements have been made on two shale materials, the Whitby mudrock and the Mancos shale (an interbedded silt and mudstone), as well as Pennant sandstone, an isotropic baseline and tight-gas sandstone analogue. Mechanical characterization includes standard triaxial experiments, pressure-dependent permeability, brazilian disk tensile strength, and fracture toughness determined using double-torsion experiments. Elastic characterisation was performed through ultrasonic velocities determined using a cross-correlation method. Additionally, we report the results of laboratory-scale fluid injection experiments for the same materials. Injection experiments involved the pressurisation of a blind-ending central hole in a dry cylindrical sample. Pressurisation is conducted under constant volume-rate control, using silicon oils of varying viscosities. Breakdown pressure is not seen to exhibit a strong dependence on rock type or orientation, and increases linearly with confining pressure. In most experiments, a small drop in the injection pressure record is observed at what is taken to be fracture initiation, and in the Pennant sandstone this is accompanied by a small burst of acoustic energy. The shale materials were acoustically quiet. Breakdown is found to be rapid and uncontrollable after initiation if injection is continued. A simplified 2-dimensional model for explaining this is presented in terms of the stress intensities at the tip of a pressurised crack, and is used alongside the triaxial data to derive a characteristic flaw size from which the fractures have initiated

Theoretical study of energetic interactions between high temperature molten materials and a low temperature fluid

International Nuclear Information System (INIS)

Chen, S.H.H.

1984-01-01

Analytical models are developed to predict the hydrodynamical transients resulting from the energetic interactions between a high temperature molten material and a low temperature liquid coolant. Initially, the molten material at high temperature and pressure is separated from the low temperature fluid by a solid metal barrier. Upon contact between the molten material and solid barrier, thermal attack occurs eventually resulting in a loss of barrier integrity. Subsequently, the molten material is injected into the liquid pool resulting in energetic interactions. The analytical models integrate a wide variety of potentially mutually-interacting transport phenomena which dominate the transient process into a deterministic scheme to predict the hydrodynamic transient process into a deterministic scheme to predict the hydrodynamic transient process. The model calculations are compared with the existing experimental results to show its engineering accuracy and adequacy in predicting such energetic interactions. Two models are formulated to bracket the transport of molten material to the rupture site for the reactor system. The stratified model minimized the rate of transport of material to the break location while the dispersed model maximized such transport. These two models are applied to a reference pressure tube reactor to evaluate the pressure transients and the potential structural damages as a result of a postulated severe primary coolant blockage in a power channel

Effect of aviation fuel type and fuel injection conditions on the spray characteristics of pressure swirl and hybrid air blast fuel injectors

Science.gov (United States)

Feddema, Rick

fuels. Optical patternation data and line of sight laser diffraction data show that there is significant difference between jet fuels. Particularly at low fuel injection pressures (0.345 MPa) and cold temperatures (-40 C), the patternation data shows that the total surface area in the spray at 38.1 mm from the pressure swirl injector for the JP-10 fuel type is one-sixth the amount of the JP-8. Finally, this study compares the atomizer performance of a pressure swirl nozzle to a hybrid air blast nozzle. The total surface area for both the hybrid air blast nozzle and the pressure swirl nozzle show a similar decline in atomization performance at low fuel injection pressures and cold temperatures. However, the optical patternator radial profile data and the line of sight laser diffraction data show that the droplet size and spray distribution data are less affected by injection conditions and fuel type in the hybrid air blast nozzle, than they are in the pressure swirl nozzle. One explanation is that the aerodynamic forces associated with the swirler on the hybrid air blast nozzle control the distribution droplets in the spray. This is in contrast to the pressure swirl nozzle droplet distribution that is controlled by internal geometry and droplet ballistics.

Radiological consequence analyses of loss of coolant accidents of various break sizes of Pressurized Heavy Water Reactor

International Nuclear Information System (INIS)

Sanyasi Rao, V.V.S.; Hari Prasad, M.; Ghosh, A.K.

2010-01-01

For any advanced technology, it is essential to ensure that the consequences associated with the accident sequences arising, if any, from the operation of the plant are as low as possible and certainly below the guidelines/limits set by the regulatory bodies. Nuclear power is no exception to this. In this paper consequences of the events arising from Loss of Coolant Accident (LOCA) sequences in Pressurized Heavy Water Reactor (PHWR), are analysed. The sequences correspond to different break sizes of LOCA followed by the operation or otherwise of Emergency Core Cooling System (ECCS). Operation or otherwise of the containment safety systems has also been considered. It has been found that there are no releases to the environment when ECCS is available. The releases, when ECCS is not available, arise from the slack and the ground. The radionuclides considered include noble gases, iodine, and cesium. The hourly meteorological parameters (wind speed, wind direction, precipitation and stability category), considered for this study, correspond to those of Kakrapar site. The consequences evaluated are the thyroid dose and the bone marrow dose received by a person located at various distances from the release point. Isodose curves are generated. From these evaluations, it has been found that the doses are very low. The complementary cumulative frequency distributions (CCFD) for thyroid and bone marrow doses have also been presented for the cases analysed. (author)

Work related to increasing the exploitation and experimental possibilities of the RA reactor, 05. Independent CO2 loop for cooling the samples irradiated in the RA vertical experimental channels (I-IV), Part II, IZ-240-0379-1963, Vol. II Head of the low temperature RA reactor coolant loop

International Nuclear Information System (INIS)

Pavicevic, M.

1963-07-01

The objective of the project was to design the head of the CO 2 coolant loop for cooling the materials during irradiation in the RA reactor. Six heads of coolant loops will be placed in the RA reactor, two in the region of heavy water in the experimental channels VEK-6 and four in the graphite reflector in the channels VEK-G. maximum generated heat in the heads of the coolant loop is 10500 kcal/h and minimum generated heat is 1500 kcal/h. The loops are cooled by CO 2 gas, coolant flow is 420 kg/h, and the pressure is 4.5 atu. There is a need to design and construct the secondary coolant loop for the low temperature coolant loop. This volume includes technical specifications of the secondary CO 2 loop with instructions for construction and testing; needed calculations; specification of materials; cost estimation for materials, equipment and construction; and graphical documentation [sr

Coolant clean-up system in the primary coolant circuit for nuclear reactor

International Nuclear Information System (INIS)

Saito, Michio.

1981-01-01

Purpose: To maintain the quality of coolants at a prescribed level by distillating coolants in the primary coolant circuit for a BWR type reactor to remove impurities therefrom, taking out the condensates from the top of the distillation column and extracting impurities in a concentrated state from the bottom. Constitution: Coolant water for cooling the core is recycled by a recycling pump by way of a recycling pipeway in a reactor. The coolants extracted from an extraction pipeway connected to the recycling pipeway are fed into a distillation column, where distillation is taken place. Impurities in the coolants, that is, in-core corrosion products, fission products generated in the reactor core, etc. are separated by the distillation, concentrated and solidified in the bottom of the distillation column. While on the other hand, condensates removed with the impurities, that is, coolants cleaned-up are recycled to the coolant water for cooling the reactor core. (Moriyama, K.)

Comparative assessment of thermophysical and thermohydraulic characteristics of lead, lead-bismuth and sodium coolants for fast reactors

International Nuclear Information System (INIS)

2002-06-01

All prototype, demonstration and commercial liquid metal cooled fast reactors (LMFRs) have used liquid sodium as a coolant. Sodium cooled systems, operating at low pressure, are characterised by very large thermal margins relative to the coolant boiling temperature and a very low structural material corrosion rate. In spite of the negligible thermal energy stored in the liquid sodium available for release in case of leakage, there is some safety concern because of its chemical reactivity with respect to air and water. Lead, lead-bismuth or other alloys of lead, appear to eliminate these concerns because the chemical reactivity of these coolants with respect to air and water is very low. Some experts believe that conceptually, these systems could be attractive if high corrosion activity inherent in lead, long term materials compatibility and other problems will be resolved. Extensive research and development work is required to meet this goal. Preliminary studies on lead-bismuth and lead cooled reactors and ADS (accelerator driven systems) have been initiated in France, Japan, the United States of America, Italy, and other countries. Considerable experience has been gained in the Russian Federation in the course of development and operation of reactors cooled with lead-bismuth eutectic, in particular, propulsion reactors. Studies on lead cooled fast reactors are also under way in this country. The need to exchange information on alternative fast reactor coolants was a major consideration in the recommendation by the Technical Working Group on Fast Reactors (TWGFRs) to collect, review and document the information on lead and lead-bismuth alloy coolants: technology, thermohydraulics, physical and chemical properties, as well as to make an assessment and comparison with respective sodium characteristics

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Bulk coolant cavitation in LMFBR containment loading following a whole-core explosion

International Nuclear Information System (INIS)

Jones, A.V.

1977-01-01

An LMFBR core undergoing an explosion transmits energy to the containment in a series of pressure waves and the containment loading is determined by their cumulative effect. These pressure waves are modified by their interaction with the coolant through which they propagate. It is necessary to model both the induction of bulk cavitation by tension waves and the interaction of pressure waves with cavitated liquid in realistic containment loading calculations. This paper sets out the progress which has been achieved in such modelling and first indications for the effect of bulk coolant cavitation in LMFBR containment loading. Conclusions may be briefly summarised: 1) Bulk cavitation must be included in realistic containment loading calculations. 2) Phenomenological models of cavitated liquid without memory are inappropriate. The best approach is to model bubble dynamics directly, including at least momentum conservation and surface tension. 3) The containment loading resulting from a given explosion is sensitive to the state of preparation of the coolant. The number density of nucleation sites should therfore accompany the results of model tests. (Auth.)

PUMP: analog-hybrid reactor coolant hydraulic transient model

International Nuclear Information System (INIS)

Grandia, M.R.

1976-03-01

The PUMP hybrid computer code simulates flow and pressure distribution; it is used to determine real time response to starting and tripping all combinations of PWR reactor coolant pumps in a closed, pressurized, four-pump, two-loop primary system. The simulation includes the description of flow, pressure, speed, and torque relationships derived through pump affinity laws and from vendor-supplied pump zone maps to describe pump dynamic characteristics. The program affords great flexibility in the type of transients that can be simulated

Effect of spacer grid mixing vanes on coolant outlet temperature distribution

Energy Technology Data Exchange (ETDEWEB)

Raemae, Tommi; Lahtinen, Tuukka; Brandt, Tellervo; Toppila, Timo [Fortum Power and Heat, Fortum (Finland). Nuclear Competence Center

2012-08-15

In Loviisa VVER-440-type NPP the coolant outlet temperature of the hot subchannel is constantly monitored during the operation. According to the authority requirement the maximum subchannel outlet temperature must not exceed the saturation temperature. Coolant temperature distribution inside the fuel assembly is affected by the efficiency of the coolant mixing. In order to enhance the coolant mixing the fuel manufacturer is introducing the additional mixing vanes on the fuel bundle spacer grids. In the paper the effect of the different mixing vane modifications is studied with computational fluid dynamics (CFD) simulation. Goal of the modelling is to find vane modifications with which sufficient mixing is reached with acceptable increase in the spacer grid pressure loss. The results of the studies are discussed in the paper. (orig.)

Sound velocity in the coolant of boiling nuclear reactors

International Nuclear Information System (INIS)

Proskuryakov, K.N.; Parshin, D.A.; Novikov, K.S.; Galivec, E.Yu.

2009-01-01

To prevent resonant interaction between acoustic resonance and natural frequencies of FE, FA and RI oscillations, it is necessary to determine the value of EACPO. Based on results of calculations of EACPO and natural frequencies of FR, FA and RI oscillations values, it would be possible to reveal the dynamical loadings on metal that are dangerous for the initiation of cracking process in the early stage of negative condition appearance. To calculate EACPO it is necessary to know the Speed Velocity in Coolant. Now we do not have any data about real values of such important parameter as pressure pulsations propagation velocity in two phase environments, especially in conditions with variations of steam content along the length of FR, with taking into account the type of local resistances, flow geometry etc. While areas of resonant interaction of the single-phase liquid coolant with equipment and internals vibrations are estimated well enough, similar estimations in the conditions of presence of a gas and steam phase in the liquid coolant are inconvenient till now. Paper presents results of calculation of velocity of pressure pulsations distribution in two-phase flow formed in core of RBMK-1000 reactors. Feature of the developed techniques is that not only thermodynamic factors and effect of a speed difference between water and steam in a two phase flow but also geometrical features of core, local resistance, non heterogeneity in the two phase environment and power level of a reactor are considered. Obtained results evidence noticeable decreasing of velocity propagation of pressure pulsations in the presence of steam actions in the liquids. Such estimations for real RC of boiling nuclear reactors with steam-liquid coolant are obtained for the first time. (author)

Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

Energy Technology Data Exchange (ETDEWEB)

Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

Flooding of a large, passive, pressure-tube light water reactor

International Nuclear Information System (INIS)

Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

1997-01-01

A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. (orig.)

Boron, arsenic and phosphorus dopant incorporation during low temperature low pressure silicon epitaxial growth

International Nuclear Information System (INIS)

Borland, J.O.; Thompson, T.; Tagle, V.; Benzing, W.

1987-01-01

Submicron silicon epitaxial structures with very abrupt epi/substrate transition widths have been realized through the use of low temperature silicon epitaxial growth techniques. At these low temperature and low pressure epitaxial growth conditions there is minimal, if any, dopant diffusion from the substrate into the epilayer during deposition. The reincorporation of autodoped dopant as well as the incorporation of intentional dopant can be a trade-off at low temperatures and low pressures. For advanced CMOS and Bi-CMOS technologies, five to six orders of magnitude change in concentration levels are desirable. In this investigation, all of the epitaxial depositions were carried out in an AMC-7810 epi-reactor with standard jets for a turbulent mixing system, and using a modified center inject configuration to achieve a single pass laminar flow system. To simulate the reincorporation of various autodoped dopant, the authors ran a controlled dopant flow of 100 sccm for each of the three dopants (boron, phosphorus and arsenic) to achieve the controlled background dopant level in the reactor gas stream

Picosecond ballistic imaging of diesel injection in high-temperature and high-pressure air

Science.gov (United States)

Duran, Sean P.; Porter, Jason M.; Parker, Terence E.

2015-04-01

The first successful demonstration of picosecond ballistic imaging using a 15-ps-pulse-duration laser in diesel sprays at temperature and pressure is reported. This technique uses an optical Kerr effect shutter constructed from a CS2 liquid cell and a 15-ps pulse at 532 nm. The optical shutter can be adjusted to produce effective imaging pulses between 7 and 16 ps. This technique is used to image the near-orifice region (first 3 mm) of diesel sprays from a high-pressure single-hole fuel injector. Ballistic imaging of dodecane and methyl oleate sprays injected into ambient air and diesel injection at preignition engine-like conditions are reported. Dodecane was injected into air heated to 600 °C and pressurized to 20 atm. The resulting images of the near-orifice region at these conditions reveal dramatic shedding of the liquid near the nozzle, an effect that has been predicted, but to our knowledge never before imaged. These shedding structures have an approximate spatial frequency of 10 mm-1 with lengths from 50 to 200 μm. Several parameters are explored including injection pressure, liquid fuel temperature, air temperature and pressure, and fuel type. Resulting trends are summarized with accompanying images.

Measuring device for the coolant flowrate in a reactor core

International Nuclear Information System (INIS)

Sawa, Toshihiko.

1983-01-01

Purpose: To improve the operation performance by enabling direct and accurate measurement for the reactor core recycling flowrate. Constitution: A control rod guide is disposed to the upper end of a control rod drive mechanism housing passing through the bottom of a reactor pressure vessel and it is inserted into the through hole of a reactor core support plate. A water flow passage is formed through the reactor core support plate for the flowrate measurement of coolants recycled within the reactor core. The static pressure difference between the upper and the lower sides of the reactor core support plate is measured by a pressure difference detector of a pressure difference measuring mechanism, and an output signal from the pressure different detector is inputted to a calculation means, in which the amount of the coolants passing through the water flow passage is calculated based on the output signal corresponding to the pressure difference. Then, the total recycling flowrate in the reactor core is determined in the calculation means based on the relation between the measured flowrate and a predetermined total reactor core recycling flowrate. (Horiuchi, T.)

MDEP Technical Report TR-CSWG-03. Technical Report: fundamental attributes for the design and construction of reactor coolant pressure-boundary components

International Nuclear Information System (INIS)

2014-01-01

The primary, long-term goal of MDEP's CSWG is to achieve international harmonisation of codes and standards for pressure boundary components in nuclear power plants that are important to reactor safety. The key to achieving harmonisation is to understand the extent of similarities and differences amongst the pressure boundary codes and standards used in various countries. To assist the CSWG in its long-term goals, several standards development organisations (SDOs) from various countries performed a comparison of their pressure boundary codes and standards to identify the extent of similarities and differences in code requirements and the reasons for their differences. This CSWG document provides the fundamental attributes which have been developed for the codes and standards used in the design and construction of reactor coolant pressure boundary components in nuclear power plants. The fundamental attributes are the basic concepts to be considered in the design, materials, fabrication, installation, examination, testing and over-pressure protection requirements for pressure boundary components

Design on Hygrometry System of Primary Coolant Circuit of HTR-PM

International Nuclear Information System (INIS)

Sun Yanfei; Zhong Shuoping; Huang Xiaojin

2014-01-01

Helium is the primary coolant in HTR-PM. If vapor get into the helium in primary coolant circuit because of some special reasons, such as the broken of steam-generator tube, chemical reaction will take effect between the graphite in reactor core and vapor in primary coolant circuit, and the safety of the reactor operation will be influenced. So the humidity of the helium in primary coolant circuit is one key parameter of HTR-PM to be monitored in-line. Once the humidity is too high, trigger signal of turning off the reactor must be issued. The hygrometry system of HTR-PM is consisting of filter, cooler, hygrometry sensor, flow meter, and some valves and tube. Helium with temperature of 250℃ is lead into the hygrometry system from the outlet of the main helium blower. After measuring, the helium is re-injected back to the primary circuit. No helium loses in this processing, and no other pump is needed. Key factors and calculations in design on hygrometry system of HTR-PM are described. A sample instrument has been made. Results of experiments proves that this hygrometry system is suitable for monitoring the humidity of the primary coolant of HTR-PM. (author)

High temperature and high performance light water cooled reactors operating at supercritical pressure, research and development

International Nuclear Information System (INIS)

Oka, Y.; Koshizuka, S.; Katsumura, Y.; Yamada, K.; Shiga, S.; Moriya, K.; Yoshida, S.; Takahashi, H.

2003-01-01

The concept of supercritical-pressure, once-through coolant cycle nuclear power plant (SCR) was developed at the University of Tokyo. The research and development (R and D) started worldwide. This paper summarized the conceptual design and R and D in Japan. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical fossil fired power plants (FPP) in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil fired power plants will be fully utilized for SCR. The high temperature, supercritical-pressure light water reactor is the logical evolution of LWR. Boiling evolved from circular boilers, water tube boilers and once-through boilers. It is the reactor version of the once-through boiler. The development from LWR to SCR follows the history of boilers. The goal of the R and D should be the capital cost reduction that cannot be achieved by the improvement of LWR. The reactor can be used for hydrogen production either by catalysis and chemical decomposition of low quality hydrocarbons in supercritical water. The reactor is compatible with tight lattice fast core for breeders due to low outlet coolant density, small coolant flow rate and high head coolant pumps

Water cooled static pressure probe

Science.gov (United States)

Lagen, Nicholas T. (Inventor); Eves, John W. (Inventor); Reece, Garland D. (Inventor); Geissinger, Steve L. (Inventor)

1991-01-01

An improved static pressure probe containing a water cooling mechanism is disclosed. This probe has a hollow interior containing a central coolant tube and multiple individual pressure measurement tubes connected to holes placed on the exterior. Coolant from the central tube symmetrically immerses the interior of the probe, allowing it to sustain high temperature (in the region of 2500 F) supersonic jet flow indefinitely, while still recording accurate pressure data. The coolant exits the probe body by way of a reservoir attached to the aft of the probe. The pressure measurement tubes are joined to a single, larger manifold in the reservoir. This manifold is attached to a pressure transducer that records the average static pressure.

Sector Tests of a Low-NO(sub x), Lean, Direct- Injection, Multipoint Integrated Module Combustor Concept Conducted

Science.gov (United States)

Tacina, Robert R.; Wey, Chang-Lie; Laing, Peter; Mansour, Adel

2002-01-01

The low-emissions combustor development described is directed toward advanced high pressure aircraft gas-turbine applications. The emphasis of this research is to reduce nitrogen oxides (NOx) at high-power conditions and to maintain carbon monoxide and unburned hydrocarbons at their current low levels at low power conditions. Low-NOx combustors can be classified into rich-burn and lean-burn concepts. Lean-burn combustors can be further classified into lean-premixed-prevaporized (LPP) and lean direct injection (LDI) concepts. In both concepts, all the combustor air, except for liner cooling flow, enters through the combustor dome so that the combustion occurs at the lowest possible flame temperature. The LPP concept has been shown to have the lowest NOx emissions, but for advanced high-pressure-ratio engines, the possibility of autoignition or flashback precludes its use. LDI differs from LPP in that the fuel is injected directly into the flame zone, and thus, it does not have the potential for autoignition or flashback and should have greater stability. However, since it is not premixed and prevaporized, good atomization is necessary and the fuel must be mixed quickly and uniformly so that flame temperatures are low and NOx formation levels are comparable to those of LPP. The LDI concept described is a multipoint fuel injection/multiburning zone concept. Each of the multiple fuel injectors has an air swirler associated with it to provide quick mixing and a small recirculation zone for burning. The multipoint fuel injection provides quick, uniform mixing and the small multiburning zones provide for reduced burning residence time, resulting in low NOx formation. An integrated-module approach was used for the construction where chemically etched laminates, diffusion bonded together, combine the fuel injectors, air swirlers, and fuel manifold into a single element. The multipoint concept combustor was demonstrated in a 15 sector test. The configuration tested had 36

Optimization of injection pressure for a compression ignition engine ...

African Journals Online (AJOL)

user

injection and atomization and contributes to incomplete combustion, nozzle clogging, ... this non edible oil may be an appropriate substitute for diesel fuel. ... The effect of injector opening pressure on the performance of a jatropha oil fuelled ...

Recent results from the MIT in-core experiments on coolant chemistry

International Nuclear Information System (INIS)

Harling, O.K.; Kohse, G.E.; Cabello, E.C.; Bernard, J.A.

1993-01-01

This paper reports results from an ongoing series of in-core experiments that have been conducted at the 5-MW(thermal) MIT Research Reactor (MITR-II) for optimizing coolant chemistries in light water reactors. Four experiments are in progress, including a pressurized coolant chemistry loop (PCCL), a boiling coolant chemistry loop (BCCL), a facility for the study of irradiation-assisted stress-corrosion cracking, and one for the evaluation of in situ sensors for the monitoring of crack propagation in metal (SENSOR). The first two have now been fully operational for several years. The latter two are scheduled to begin regular operation later this year

AMPTRACT: an algebraic model for computing pressure tube circumferential and steam temperature transients under stratified channel coolant conditions

International Nuclear Information System (INIS)

Gulshani, P.; So, C.B.

1986-10-01

In a number of postulated accident scenarios in a CANDU reactor, some of the horizontal fuel channels are predicted to experience periods of stratified channel coolant condition which can lead to a circumferential temperature gradient around the pressure tube. To study pressure tube strain and integrity under stratified flow channel conditions, it is, necessary to determine the pressure tube circumferential temperature distribution. This paper presents an algebraic model, called AMPTRACT (Algebraic Model for Pressure Tube TRAnsient Circumferential Temperature), developed to give the transient temperature distribution in a closed form. AMPTRACT models the following modes of heat transfer: radiation from the outermost elements to the pressure tube and from the pressure to calandria tube, convection between the fuel elements and the pressure tube and superheated steam, and circumferential conduction from the exposed to submerged part of the pressure tube. An iterative procedure is used to solve the mass and energy equations in closed form for axial steam and fuel-sheath transient temperature distributions. The one-dimensional conduction equation is then solved to obtain the pressure tube circumferential transient temperature distribution in a cosine series expansion. In the limit of large times and in the absence of convection and radiation to the calandria tube, the predicted pressure tube temperature distribution reduces identically to a parabolic profile. In this limit, however, radiation cannot be ignored because the temperatures are generally high. Convection and radiation tend to flatten the parabolic distribution

Effect of injection pressure on heat release rate and emissions in CI engine using orange skin powder diesel solution

International Nuclear Information System (INIS)

Purushothaman, K.; Nagarajan, G.

2009-01-01

Experiments have been conducted to study the effect of injection pressure on the combustion process and exhaust emissions of a direct injection diesel engine fueled with Orange Skin Powder Diesel Solution (OSPDS). Earlier investigation by the authors revealed that 30% OSPDS was optimum for better performance and emissions. In the present investigation the injection pressure was varied with 30% OSPDS and the combustion, performance and emissions characteristics were compared with those of diesel fuel. The different injection pressures studied were 215 bar, 235 bar and 255 bar. The results showed that the cylinder pressure with 30% OSPDS at 235 bar fuel injection pressure, was higher than that of diesel fuel as well as at other injection pressures. Similarly, the ignition delay was longer and with shorter combustion duration with 30% OSPDS at 235 bar injection pressure. The brake thermal efficiency was better at 235 bar than that of other fuel injection pressures with OSPDS and lower than that of diesel fuel. The NO x emission with 30% OSPDS was higher at 235 bar. The hydrocarbon and CO emissions were lower with 30% OSPDS at 235 bar. The smoke emission with 30% OSPDS was marginally lower at 235 bar and marginally higher at 215 bar than for diesel fuel. The combustion, performance and emission characteristics of the engine operating on the test fuels at 235 bar injection pressure were better than other injection pressures

An analysis of system pressure and temperature distribution in self-pressurizer of SMART and calculation of sizing of wet thermal insulator and pressurizer cooler

Energy Technology Data Exchange (ETDEWEB)

Kang, Yeon Moon; Lee, Doo Jeong; Yoon, Ju Hyun; Kim, Hwan Yeol [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-03-01

To evaluate the amount of heat transfer from coolant to gas in reactor vessel heat transfer through the structure of pressurizer and evaporation/condensation on surface of liquid pool should be considered. And, also the heat exchange by pressurizer cooler and heat transfer to upper plate of reactor vessel should be considered. Thus, overall examinations on design variables which affect the heat transfer from coolant to gas are needed to maintain the pressurizer conditions at designed value for normal operation through heatup process. The major design variables, which affect system pressure and gas temperature during heatup, and the sizes of wet thermal insulator and pressurizer cooler, and volume of gas cylinder connected to pressurizer. A computer program is developed for the prediction of system pressure and temperature of pressurizer gas region with considering volume expansion of coolant and heat transfer from coolant to gas during heatup. Using the program, this report suggests the optimized design values of wet thermal insulator, pressurizer cooler, and volume of gas cylinder to meet the target conditions for normal operation of SMART. (author). 6 refs., 17 figs., 5 tabs.

Study on enhanced lymphatic tracing of isosulfan blue injection by influence of osmotic pressure on lymphatic exposure.

Science.gov (United States)

Ye, Tiantian; He, Rui; Wu, Yue; Shang, Lei; Wang, Shujun

2018-04-01

Isosulfan blue (IB) is being used as a lymphatic tracer has been approved by the FDA in 1981. This study aimed at improving lymphatic exposure of IB injection by osmotic pressure regulation to achieve step-by step lymphatic tracing. First, IB injection with appropriate osmotic pressure, stability, and suitable pH was prepared. Next, the lymphatic tracing ability of different osmotic pressure was studied to determine the blue-stained state of IB in three-level lymph nodes after subcutaneous administration. Furthermore, pharmacokinetics of lymphatic drainage, lymph node uptake, and plasma concentration was investigate to explore the improving law of the lymphatic tracing by osmotic pressure, and combined with tissue irritation to determine the optimal osmotic pressure. At last, the tissue distribution in mice of IB injection which had the property of optimal osmotic pressure was investigated. The results showed that increasing osmotic pressure could significantly reduce injection site retention and increase IB concentration of lymph node. The lymph nodes could be obviously blue-stained by IB injection which had 938 mmol/kg osmotic pressure and would not cause inflammatory reaction and blood exposure. The tissue distribution study suggested that IB injection which had 938 mmol/kg osmotic pressure was mainly distributed into gallbladder and duodenum that verified the reports that 90% IB was excreted through the feces through biliary excretion. In conclusion, this study provides the basic study to improve lymphatic exposure of IB injection by regulate the osmotic pressure and have the potential to be the helpful guidance for the elective lymph node dissection.

Radiolysis of the VVER-1000 reactor coolant: An experimental study and mathematical modeling

International Nuclear Information System (INIS)

Arkhipov, O.P.; Bugaenko, V.L.; Kabakchi, S.A.

1995-01-01

Variations in the composition of the coolant for the primary circuit of a VVER-1000 reactor of the Kalinin nuclear power plant upon transition from power-level operation to shutdown was studied experimentally. The data obtained were used for verification of the MORAVA-H2 program developed earlier for simulation of the coolant state in pressurized-water power reactors

Universal treatment of plumes and stresses for pressurized thermal shock evaluations

International Nuclear Information System (INIS)

Theofanous, T.G.; Angelini, S.; Yan, H.

1991-01-01

Thermally-induced stresses in a reactor pressure vessel wall, as a result of high-pressure safety injection, are an essential component of integrated risk analyses of pressurized thermal shock transients. Limiting cooldowns arise when this injection occurs under stagnated loop conditions which, in turn, correspond to a rather narrow range (in size) of small-break loss-of-coolant accidents. Moreover, at these conditions, the flow is thermally stratified, and in addition to the global cooldown, one must be concerned about the additional cooling potential due to the downcomer plumes formed by the cold streams pouring out of the cold legs. In the Nuclear Regulatory Commission's Integrated Pressurized Thermal Shock (IPTS) study, this stratification was calculated with the codes REMIX/NEWMIX. A comprehensive comparison with all available experimental data has currently been compiled. The stress analysis using this input was carried out at Oak Ridge National Laboratory using a one-dimensional approximation with the intent to conservatively bound the magnitude of thermal stresses

Evaluation of stress histories of reactor coolant loop piping for pipe rupture prediction

International Nuclear Information System (INIS)

Lu, S.C.; Larder, R.A.; Ma, S.M.

1981-01-01

This paper describes the analyses used to evaluate stress histories in the primary coolant loop piping of a selected four-loop PNR power station. In order to make the simulation as realistic as possible, best estimates rather than conservative assumptions were considered throughout. The best estimate solution, however, was aided by a sensitivity study to assess the possible variation of outcomes resulted from uncertainties associated with these assumptions. Sources of stresses considered in the evaluation were pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, pump vibrations, and finally seismic excitations. The best estimates of pressure and thermal transient histories arising from plant operations were based on actual plant operation records supplemented by specified plant design conditions. Seismic motions were generated from response spectrum curves developed specifically for the region surrounding the plant site. Stresses due to dead weight and thermal expansion were computed from a three dimensional finite element model which used a combination of pipe, truss, and beam elements to represent the coolant loop piping, the pressure vessel, coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients were obtained by closed form solutions. Seismic stress calculations considered the soil structure interaction, the coupling effect between the containment structure and the reactor coolant system. A time history method was employed for the seismic analysis. Calculations of residual stresses accounted for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation were estimated by a dynamic analysis using existing measurements of pump vibrations. (orig./HP)

Transient heat transfer phenomena of the liquid metal layer cooled by overlying R113 coolant

International Nuclear Information System (INIS)

Cho, J. S.; Seo, K. R.; Jung, C. H.; Park, R. J.; Kim, S. B.

1999-01-01

To understand the fundamental relationship of the natural convection heat transfer in the molten metal pool and the boiling mechanism of the overlying coolant, experiments were performed for the transient heat transfer of the liquid metal pool with overlying R113 coolant with boiling. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. The metal pool is heated from the bottom surface and the coolant is injected onto the molten metal pool. Tests were conducted by changing the bottom surface boundary condition. The bottom heating condition was varied from 8kW to 14kW. As a result the boiling mechanism of the R113 coolant is changed from the nuclear boiling to film boiling. The Nusselt number and the Rayleigh number in the molten metal pool region obtained as functions of time. Analysis was made for the relationship between the heat flux and the temperature difference of the metal layer surface temperature and the boiling coolant bulk temperature

SMART core power control method by coolant temperature variation

International Nuclear Information System (INIS)

Lee, Chung Chan; Cho, Byung Oh

2001-08-01

SMART is a soluble boron-free integral type pressurized water reactor. Its moderator temperature coefficient (MTC) is strongly negative throughout the cycle. The purpose of this report is how to utilize the primary coolant temperature as a second reactivity control system using the strong negative MTC. The reactivity components associated with reactor power change are Doppler reactivity due to fuel temperature change, moderator temperature reactivity and xenon reactivity. Doppler reactivity and moderator temperature reactivity take effects almost as soon as reactor power changes. On the other hand, xenon reactivity change takes more than several hours to reach an equilibrium state. Therefore, coolant temperature at equilibrium state is chosen as the reference temperature. The power dependent reference temperature line is limited above 50% power not to affect adversely in reactor safety. To compensate transient xenon reactivity, coolant temperature operating range is expanded. The suggested coolant temperature operation range requires minimum control rod motion for 50% power change. For smaller power changes such as 25% power change, it is not necessary to move control rods to assure that fuel design limits are not exceeded

Numerical Simulation of the Pressure Distribution in the Reactor Vessel Downcomer Region Fluctuated by the Reactor Coolant Pump

Energy Technology Data Exchange (ETDEWEB)

Jeong, Dong Hwa; Jung, Byung Ryul; Jang, Ho Cheol; Yune, Seok Jeong; Kim, Eun Kee [KEPCO EnC, Daejeon (Korea, Republic of)

2015-10-15

In this study the numerical simulation of the pressure distribution in the downcomer region resulting from the pressure pulsation by the Reactor Coolant Pump (RCP) is performed using the Finite Difference Method (FDM). Simulation is carried out for the cylindrical shaped 2-dimensional model equivalent to the outer surface of the Core Support Barrel (CSB) of APR1400 and a 1/2 model is adopted based on the bilateral symmetry by the inlet nozzle. The fluid temperature is 555 .deg. F and the forcing frequencies are 120Hz, 240Hz, 360Hz and 480Hz. Simulation results of the axial pressure distributions are provided as the Root Mean Square (RMS) values at the five locations of 0°, 45°, 90°, 135° and 180° in the circumferential direction from the inlet nozzle location. In the study, the numerical simulation of pressure distributions in the downcomer region induced by the RCP was performed using FDM and the results were reviewed. The interference of the waves returned from both boundaries in the axial direction and the source of the sinusoidal wave is shown on the inlet nozzle interface pressure point. It seems that the maximum pressures result from the superposition of the waves reflected from the seating surface and the waves newly arrived from the inlet nozzle interface pressure location.

Investigation of High Pressure, Multi-Hole Diesel Fuel Injection Using High Speed Imaging

Science.gov (United States)

Morris, Steven; Eagle, Ethan; Wooldridge, Margaret

2012-10-01

Research to experimentally capture and understand transient fuel spray behavior of modern fuel injection systems remains underdeveloped. To this end, a high-pressure diesel common-rail fuel injector was instrumented in a spherical, constant volume combustion chamber to image the early time history of injection of diesel fuel. The research-geometry fuel injector has four holes aligned on a radial plane of the nozzle with hole sizes of 90, 110, 130 and 150 μm in diameter. Fuel was injected into a non-reacting environment with ambient densities of 17.4, 24.0, and 31.8 kg/m3 at fuel rail pressures of 1000, 1500, and 2000 bar. High speed images of fuel injection were taken using backlighting at 100,000 frames per second (100 kfps) and an image processing algorithm. The experimental results are compared with a one-dimensional fuel-spray model that was historically developed and applied to fuel sprays from single-hole fuel injectors. Fuel spray penetration distance was evaluated as a function of time for the different injector hole diameters, fuel injection pressures and ambient densities. The results show the differences in model predictions and experimental data at early times in the spray development.

Correlation of cylinder-head temperatures and coolant heat rejections of a multicylinder, liquid-cooled engine of 1710-cubic-inch displacement

Science.gov (United States)

Lundin, Bruce T; Povolny, John H; Chelko, Louis J

1949-01-01

Data obtained from an extensive investigation of the cooling characteristics of four multicylinder, liquid-cooled engines have been analyzed and a correlation of both the cylinder-head temperatures and the coolant heat rejections with the primary engine and coolant variables was obtained. The method of correlation was previously developed by the NACA from an analysis of the cooling processes involved in a liquid-cooled-engine cylinder and is based on the theory of nonboiling, forced-convection heat transfer. The data correlated included engine power outputs from 275 to 1860 brake horsepower; coolant flows from 50 to 320 gallons per minute; coolants varying in composition from 100 percent water to 97 percent ethylene glycol and 3 percent water; and ranges of engine speed, manifold pressure, carburetor-air temperature, fuel-air ratio, exhaust-gas pressure, ignition timing, and coolant temperature. The effect on engine cooling of scale formation on the coolant passages of the engine and of boiling of the coolant under various operating conditions is also discussed.

Responses to Small Break Loss of Coolant Accidents for SMART

International Nuclear Information System (INIS)

Bae, Kyoo Hwan; Kim, Hee C.; Chang, Moon H.; Zee, Sung Q.; Kim, Si-Hwan; Lee, Un-Chul

2004-01-01

The SMART NSSS adopts the design characteristics of containing most of the primary circuit components, such as the reactor core, main coolant pumps (MCPs), steam generators (SGs), and N 2 gas pressurizer (PZR) in a single leak-tight Reactor Pressure Vessel (RPV) with a relatively large ratio of the primary coolant inventory to the core power compared to the conventional loop-type PWR. Due to these design characteristics, the SMART can fundamentally eliminate the possibility of Large Break Loss of Coolant Accidents (LBLOCAs), improve the natural circulation capability, and assure a sufficient time to mitigate the possibility of core uncover. Also, SMART adopts inherent safety improving features and passive engineered safety systems such as the substantially large negative moderator temperature coefficients, passive residual heat removal system, emergency core cooling system, and a steel-made leak-tight Safeguard Vessel (SV) housing the RPV. This paper presents the results of the safety analyses using a MARS/SMR code for the instantaneous guillotine ruptures of the major pipelines penetrating the RPV. The analysis results, employing conservative initial/boundary conditions and assumptions, show that the safety systems of the SMART basic design adequately remove the core decay heat without causing core uncover for all the cases of the Small Break Loss of Coolant Accidents (SBLOCAs). The sensitivity study results with variable SV conditions show that the reduced SV net free volume can shorten the time for reaching the thermal and mechanical equilibrium condition between the RPV and SV. Under these boundary conditions, the primary system inventory loss can be minimized and the core remains covered for a longer period of time without any makeup of the coolant. (authors)

Validation of Computational Fluid Dynamics Calculation Using Rossendorf Coolant Mixing Model Flow Measurements in Primary Loop of Coolant in a Pressurized Water Reactor Model

Directory of Open Access Journals (Sweden)

Istvan Farkas

2016-08-01

Full Text Available The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively with experimental results.

Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

International Nuclear Information System (INIS)

Forsberg, C.W.

2002-01-01

The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

Coolant clean-up system in nuclear reactor