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

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.)

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

Coolant clean up system in nuclear reactor

International Nuclear Information System (INIS)

Tajima, Fumio; Iwami, Hiroshi.

1981-01-01

Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

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

2003-01-01

To support the development of a Probabilistic Safety Assessment (PSA) model usable in Riskinformed Applications (RIA) for Korea Standard Nuclear power Plants (KSNP), we have performed a thermal hydraulic analysis of Aggressive Secondary Cooldown (ASC) in a 2-inch Small Break Loss Of Coolant Accident (SBLOCA) with a total loss of High Pressure Safety Injection (HPSI). The present study focuses on the estimation of the success criteria of ASC, and the enhanced understanding of the detailed thermal hydraulic behavior and phenomena. The results have shown that the Reactor Coolant System (RCS) pressure can be reduced to the Low Pressure Safety Injection (LPSI) operation conditions without core damage. It was also shown that more relaxed success criteria compared to those in the previous PSA models of KSNP could be used in the new PSA model. However, it was found that the results could be affected by various parameters related with ASC operation, i.e., reference temperature for the calculation of the cooldown rate and its control method

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

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)

Performance Analysis of Multi Stage Safety Injection Tank

International Nuclear Information System (INIS)

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

2015-01-01

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

Performance Analysis of Multi Stage Safety Injection Tank

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Coolant make-up device for BWR type reactor

International Nuclear Information System (INIS)

Sasagawa, Hiroshi.

1994-01-01

In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

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.)

ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project

Science.gov (United States)

Morrison, Russell H.; Holt, Mike

2005-01-01

The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

Full reactor coolant system chemical decontamination qualification programs

Energy Technology Data Exchange (ETDEWEB)

Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1995-03-01

Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

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

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

Liquid metal coolants for fusion-fission hybrid system: A neutronic analysis

Energy Technology Data Exchange (ETDEWEB)

Marques, Renato V.A.; Velasquez, Carlos E.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L., E-mail: claubia@nuclear.ufmg.br [Universidade de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Barros, Graiciany P. [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

Based on a work already published by the UFMG Nuclear Engineering Department, it was suggested to use different coolant materials in a fusion-fission system after a fuel burnup simulation, including that one used in reference work. The goal is to compare the neutron parameters, such as the effect multiplication factor and actinide amounts in transmutation layer, for each used coolant and find the best(s) coolant material(s) to be applied in the considered system. Results indicate that the lead and lead-bismuth coolant are the most suitable choices to be applied to cool the system. (author)

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)

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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.

System approach in the investigation of coolant parametrical oscillations in passive safety injection systems (PSIS)

International Nuclear Information System (INIS)

Proskouriakov, K.N.

2001-01-01

The use of thermal-hydraulic computer codes is an important part of the work programme for activities in the field of nuclear power plants (NPP) Safety Research as it will enable to define better the test configuration and parameter range extensions and to extrapolate the results of the small scale experiments towards full scale reactor applications. The CATHARE2, RELAP5, the WCOBRA/TRAC, and APROS codes are the estimate thermal hydraulic codes for the evaluation of large and small break loss of coolant accidents (LOCA). The relatively good agreement experimental data with the calculations have been presented. There was shown also some big mistakes in predicting distribution of flow when two phase are present. Model of parametrical oscillation (P.O.) worked out gives explanation for flow oscillations and indicates that the phenomenon of P.O. appears under certain combination of thermal-hydraulic parameters and structure of heat-removal system. (orig.)

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.

ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project -2006 Update

Science.gov (United States)

Morrison, Russell H.; Holt, Mike

2006-01-01

The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. This paper presents a status of the coolant stability over the past year as well as results from destructive analyses of hardware removed from the on-orbit system and the current approach to coolant remediation.

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)

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)

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

Coolant controls of a PEM fuel cell system

Science.gov (United States)

Ahn, Jong-Woo; Choe, Song-Yul

When operating the polymer electrolyte membrane (PEM) fuel cell stack, temperatures in the stack continuously change as the load current varies. The temperature directly affects the rate of chemical reactions and transport of water and reactants. Elevated temperature increases the mobility of water vapor, which reduces the ohmic over-potential in the membrane and eases removal of water produced. Adversely, the high temperature might impose thermal stress on the membrane and cathode catalyst and cause degradation. Conversely, excessive supply of coolants lowers the temperature in the stack and reduces the rate of the chemical reactions and water activity. Corresponding parasitic power dissipated at the electrical coolant pump increases and overall efficiency of the power system drops. Therefore, proper design of a control for the coolant flow plays an important role in ensuring highly reliable and efficient operations of the fuel cell system. Herein, we propose a new temperature control strategy based on a thermal circuit. The proposed thermal circuit consists of a bypass valve, a radiator with a fan, a reservoir and a coolant pump, while a blower and inlet and outlet manifolds are components of the air supply system. Classic proportional and integral (PI) controllers and a state feedback control for the thermal circuit were used in the design. In addition, the heat source term, which is dependent upon the load current, was feed-forwarded to the closed loop and the temperature effects on the air flow rate were minimized. The dynamics and performance of the designed controllers were evaluated and analyzed by computer simulations using developed dynamic fuel cell system models, where a multi-step current and an experimental current profile measured at the federal urban driving schedule (FUDS) were applied. The results show that the proposed control strategy cannot only suppress a temperature rise in the catalyst layer and prevent oxygen starvation, but also reduce the

Conceptual design of the integral test loop (I): Reactor coolant system and secondary system

Energy Technology Data Exchange (ETDEWEB)

Song, Chul Hwa; Lee, Seong Je; Kwon, Tae Soon; Moon, Sang Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-10-01

This report describes the conceptual design of the primary coolant system and the secondary system of the Integral Test Loop (ITL) which simulates overall thermal hydraulic phenomena of the primary system of a nuclear power plant during postulated accidents or transients. The design basis for the primary coolant system and secondary system is as follows ; Reference plant: Korean Standard Nuclear Plant (KSNP), Height ratio : 1/1, Volume ratio : 1/200, Power scale : Max. 15% of the scaled nominal power, Temperature, Pressure : Real plant conditions. The primary coolant system includes a reactor vessel, which contains a core simulator, a steam generator, a reactor coolant pump simulator, a pressurizer and piping, which consists of two hot legs, four cold legs and four intermediate legs. The secondary system consists of s steam discharge system, a feedwater supply system and a steam condensing system. This conceptual design report describes general configuration of the reference plant, and major function and operation of each system of the plant. Also described is the design philosophy of each component and system of the ITL, and specified are the design criteria and technical specifications of each component and system of the ITL in the report. 17 refs., 43 figs., 51 tabs. (Author)

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

International Nuclear Information System (INIS)

1976-09-01

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

Performance Analysis of Thermoelectric Based Automotive Waste Heat Recovery System with Nanofluid Coolant

Directory of Open Access Journals (Sweden)

Zhi Li

2017-09-01

Full Text Available Output performance of a thermoelectric-based automotive waste heat recovery system with a nanofluid coolant is analyzed in this study. Comparison between Cu-Ethylene glycol (Cu-EG nanofluid coolant and ethylene glycol with water (EG-W coolant under equal mass flow rate indicates that Cu-EG nanofluid as a coolant can effectively improve power output and thermoelectric conversion efficiency for the system. Power output enhancement for a 3% concentration of nanofluid is 2.5–8 W (12.65–13.95% compared to EG-Water when inlet temperature of exhaust varies within 500–710 K. The increase of nanofluid concentration within a realizable range (6% has positive effect on output performance of the system. Study on the relationship between total area of thermoelectric modules (TEMs and output performance of the system indicates that optimal total area of TEMs exists for maximizing output performance of the system. Cu-EG nanofluid as coolant can decrease optimal total area of TEMs compared with EG-W, which will bring significant advantages for the optimization and arrangement of TEMs whether the system space is sufficient or not. Moreover, power output enhancement under Cu-EG nanofluid coolant is larger than that of EG-W coolant due to the increase of hot side heat transfer coefficient of TEMs.

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.

Coolant clean-up system in nuclear reactor

International Nuclear Information System (INIS)

Tsuburaya, Hirobumi; Akita, Minoru; Shiraishi, Tadashi; Kinoshita, Shoichiro; Okura, Minoru; Tsuji, Akio.

1987-01-01

Purpose: To ensure a sufficient urging pressure at the inlet of a coolant clean-up system pump in a nuclear reactor and eliminate radioactive contaminations to the pump. Constitution: Coolant clean-up system (CUW) pump in a nuclear reactor is disposed to the downstream of a filtration desalter and, for compensating the insufficiency of the urging pressure at the pump inlet, the reactor water intake port to the clean-up system is disposed to the downstream of the after-heat removing pump and the heat exchanger. By compensating the net positive suction head (NPSH) of the clean-up system from the residual heat removing system, the problems of insufficient NPSH for the CUW pump upon reactor shut-down can be dissolved and, accordingly, the reactor clean-up system can be arranged in the order of the heat exchanger, clean-up device and pump. Thus, the CUW pump acts on reactor water after cleaned-up in the clean-up device to reduce the radioactivity contamination to the pump. (Kawakami, Y.)

THE PROBLEM OF ENERGY EFFICIENCY OF THE GEOTHERMAL CIRCULATION SYSTEM IN DIFFERENT MODES OF REINJECTION OF THE COOLANT

Directory of Open Access Journals (Sweden)

D. K. Djavatov

2017-01-01

Full Text Available Aim. Advanced technologies are crucial for widespread use of geothermal energy to ensure its competitiveness with conventional forms of energy. To date, the basis for the development of geothermal energy is the technology of extracting the heat transfer fluids from the subsoil. There are the following ways to extract the coolant: freeflow; pumping and circular methods. Of greatest interest is the technology to harness the geothermal energy based on geothermal circulatory system (GCS. There is the problem of the right choice of technological parameters for geothermal systems to ensure their effective functioning.Methods. We consider the development of geothermal energy technology based on geothermal circulatory system, as this technology solves the dumping of the waste water containing environmentally harmful substances. In addition to the environmental issues, this technology makes it possible to intensify the process of production and the degree of extraction of thermal resources, which significantly increases the potential for geothermal heat resources in terms of the fuel and energy balance.Findings. Were carried out optimization calculations for Ternairsky deposits of thermal waters. In the calculations, was taken into account the temperature dependence of important characteristics, such as the density and heat capacity of the coolant.Conclusions. There is the critical temperature of the coolant injected, depending on the flow rate and the diameter of the well, ensuring the effective functioning of the geothermal circulatory systems. 

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.)

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)

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

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

Design of Reactor Coolant Pump Seal Online Monitoring System

International Nuclear Information System (INIS)

Ah, Sang Ha; Chang, Soon Heung; Lee, Song Kyu

2008-01-01

As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation

Design of Reactor Coolant Pump Seal Online Monitoring System

Energy Technology Data Exchange (ETDEWEB)

Ah, Sang Ha; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of); Lee, Song Kyu [Korea Power Engineering Co., Yongin (Korea, Republic of)

2008-05-15

As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation.

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.

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

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

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

The effect of zinc injection into PWR primary coolant on the reduction of radiation buildup and corrosion control. The solubilities of zinc, nickel and cobalt spinel oxides

International Nuclear Information System (INIS)

Miyajima, Kaori; Hirano, Hideo

1999-01-01

The use of zinc injection into PWR primary coolant to reduce radiation buildup has been widely studied, and te reduction effect has been experimentally confirmed. However, some items, such as the optimal concentration of zinc required to reduce radiation buildup, the corrosion control effect of zinc injection, and the influence of zinc injection on the integrity of fuel cladding, have not been clarified yet. In particular, the corrosion suppression effect of zinc remains unconfirmed. Therefore, it is necessary to measure and calculate the solubilities of zinc and nickel spinel oxides, which are formed on the surface of Ni-based alloys in PWR primary systems. In this study, in order to assess the effectiveness of zinc injection in the reduction of radiation buildup and the corrosion control of Ni-based alloy, the potential-pH diagrams for Zn-Cr-H 2 O, Ni-Cr-H 2 O, and Co-Cr-H 2 O systems at 300degC were constructed and the solubilities of Zn-Cr, Ni-Cr, and Co-Cr spinel oxides were calculated. It is concluded that under pH conditions for which NiCr 2 O 4 is stable, zinc injection is effective in corrosion control as well as in reducing radiation buildup. (author)

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

Coolant Design System for Liquid Propellant Aerospike Engines

Science.gov (United States)

McConnell, Miranda; Branam, Richard

2015-11-01

Liquid propellant rocket engines burn at incredibly high temperatures making it difficult to design an effective coolant system. These particular engines prove to be extremely useful by powering the rocket with a variable thrust that is ideal for space travel. When combined with aerospike engine nozzles, which provide maximum thrust efficiency, this class of rockets offers a promising future for rocketry. In order to troubleshoot the problems that high combustion chamber temperatures pose, this research took a computational approach to heat analysis. Chambers milled into the combustion chamber walls, lined by a copper cover, were tested for their efficiency in cooling the hot copper wall. Various aspect ratios and coolants were explored for the maximum wall temperature by developing our own MATLAB code. The code uses a nodal temperature analysis with conduction and convection equations and assumes no internal heat generation. This heat transfer research will show oxygen is a better coolant than water, and higher aspect ratios are less efficient at cooling. This project funded by NSF REU Grant 1358991.

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

Natural circulation in reactor coolant system

International Nuclear Information System (INIS)

Han, J.T.

1987-01-01

Reactor coolant system (RCS) natural circulation in a PWR is the buoyancy-driven coolant circulation between the core and the upper-plenum region (in-vessel circulation) with or without a countercurrent flow in the hot leg piping between the vessel and steam generators (ex-vessel circulation). This kind of multidimensional bouyancy-driven flow circulation serves as a means of transferring the heat from the core to the structures in the upper plenum, hot legs, and possibly steam generators. As a result, the RCS piping and other pressure boundaries may be heated to high temperatures at which the structural integrity is challenged. RCS natural circulation is likely to occur during the core uncovery period of the TMLB' accident in a PWR when the vessel upper plenum and hot leg are already drained and filled with steam and possibly other gaseous species. RCS natural circulation is being studied for the Surry plant during the TMLB' accident in which station blackout coincides with the loss of auxiliary feedwater and no operator actions. The effects of the multidimensional RCS natural circulation during the TMLB' accident are discussed

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.

Phenomena occuring in the reactor coolant system during severe core damage accidents

International Nuclear Information System (INIS)

Malinauskas, A.P.

1990-01-01

The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. This paper discusses, how in the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. The physical and chemical processes occurring within the RCS during normal operation of the reactor are relatively uncomplicated and are reasonably well understood. When the flow of coolant is properly adjusted, the thermal energy resulting from nuclear fission (or, in the shutdown mode, from radioactive decay processes) and secondary inputs, such as pumps, are exactly balanced by thermal losses through the RCS boundaries and to the various heat sinks that are employed to effect the conversion of heat to electrical energy. Because all of the heat and mass fluxes remain sensibly constant with time, mathematical descriptions of the thermophysical processes are relatively straightforward, even for boiling water reactor (BWR) systems. Although the coolant in a BWR does undergo phase changes, the phase boundaries remain well-defined and time-invariant

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)

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...

New cooling system of the FRG-1 two barrier system of the primary coolant cycle

International Nuclear Information System (INIS)

Knop, W.; Schreiner, P.

2003-01-01

The GKSS research center operates the swimming pool reactor FRG-1 with a thermal power of 5 MW as national neutron source for neutron scattering experiments and sample irradiation as well. Before changing the primary coolant cycle consisted of the reactor core and the closed piping including pumps, heat exchanger and delay tank. The closed cooling circuit was located underneath the reactor pool, in the so-called radioactive cellar. This piping system served secondary coolant system. Due to the location of the primary coolant cycle below the operation pool a postulated 2-F line break and simultaneous failure of the pool slide gate valve could lead to a falling dry of the total reactor core. the new primary coolant system was built in the beginning 2002 in a partitioned cell all within the radioactive cellar, so that the reactor core remains with water with the assumed incident. Due to the new two barrier-inclusion of the primary circuit only the melting of two fuel plates (from total 252 fuel plates) has to be taken into account. This measure and the core compactness in 2000 with a neutron flux gain of a factor of 2 makes the FRG-1 ready for the next 15 years of reactor operation. (author)

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

Method for controlling a coolant liquid surface of cooling system instruments in an atomic power plant

International Nuclear Information System (INIS)

Monta, Kazuo.

1974-01-01

Object: To prevent coolant inventory within a cooling system loop in an atomic power plant from being varied depending on loads thereby relieving restriction of varied speed of coolant flow rate to lowering of a liquid surface due to short in coolant. Structure: Instruments such as a superheater, an evaporator, and the like, which constitute a cooling system loop in an atomic power plant, have a plurality of free liquid surface of coolant. Portions whose liquid surface is controlled and portions whose liquid surface is varied are adjusted in cross-sectional area so that the sum total of variation in coolant inventory in an instrument such as a superheater provided with an annulus portion in the center thereof and an inner cylindrical portion and a down-comer in the side thereof comes equal to that of variation in coolant inventory in an instrument such as an evaporator similar to the superheater. which is provided with an overflow pipe in its inner cylindrical portion or down-comer, thereby minimizing variation in coolant inventory of the entire coolant due to loads thus minimizing variation in varied speed of the coolant. (Kamimura, M.)

A system for cooling electronic elements with an EHD coolant flow

International Nuclear Information System (INIS)

Tanski, M; Kocik, M; Barbucha, R; Garasz, K; Mizeraczyk, J; Kraśniewski, J; Oleksy, M; Hapka, A; Janke, W

2014-01-01

A system for cooling electronic components where the liquid coolant flow is forced with ion-drag type EHD micropumps was tested. For tests we used isopropyl alcohol as the coolant and CSD02060 diodes in TO-220 packages as cooled electronic elements. We have studied thermal characteristics of diodes cooled with EHD flow in the function of a coolant flow rate. The transient thermal impedance of the CSD02060 diode cooled with 1.5 ml/min EHD flow was 7.8°C/W. Similar transient thermal impedance can be achieved by applying to the diode a large RAD-A6405A/150 heat sink. We found out that EHD pumps can be successfully applied for cooling electronic elements.

Selection of an Alternate Biocide for the ISS Internal Thermal Control System Coolant, Phase 2

Science.gov (United States)

Wilson, Mark E.; Cole, Harold; Weir, Natalee; Oehler, Bill; Steele, John; Varsik, Jerry; Lukens, Clark

2004-01-01

The ISS (International Space Station) ITCS (Internal Thermal Control System) includes two internal coolant loops that utilize an aqueous based coolant for heat transfer. A silver salt biocide had previously been utilized as an additive in the coolant formulation to control the growth and proliferation of microorganisms within the coolant loops. Ground-based and in-flight testing demonstrated that the silver salt was rapidly depleted, and did not act as an effective long-term biocide. Efforts to select an optimal alternate biocide for the ITCS coolant application have been underway and are now in the final stages. An extensive evaluation of biocides was conducted to down-select to several candidates for test trials and was reported on previously. Criteria for that down-select included: the need for safe, non-intrusive implementation and operation in a functioning system; the ability to control existing planktonic and biofilm residing microorganisms; a negligible impact on system-wetted materials of construction; and a negligible reactivity with existing coolant additives. Candidate testing to provide data for the selection of an optimal alternate biocide is now in the final stages. That testing has included rapid biocide effectiveness screening using Biolog MT2 plates to determine minimum inhibitory concentration (amount that will inhibit visible growth of microorganisms), time kill studies to determine the exposure time required to completely eliminate organism growth, materials compatibility exposure evaluations, coolant compatibility studies, and bench-top simulated coolant testing. This paper reports the current status of the effort to select an alternate biocide for the ISS ITCS coolant. The results of various test results to select the optimal candidate are presented.

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)

In-Service Inspection system for coolant channels of Indian PHWRS - evolution and experience

International Nuclear Information System (INIS)

Puri, R.K.; Singh, M.

2006-01-01

In-Service Inspection (ISI) is the most important of all periodic monitoring and surveillance activities for assuring the structural integrity of coolant channels in the life extension and management of pressurized heavy water reactors (PHWR-CANDU). Indian PHWRs (220 MWe) are characterized by consists by 306 coolant channels in each unit. These channels have to be inspected for various parameters over the operating life of the reactor. ISI of coolant channels necessitated the indigenous development of an inspection system called BARCIS (BARC Channel Inspection System) at Bhabha Atomic Research Center. BARCIS consists of mainly three parts; drive and control unit, special sealing plug and an inspection head carrying various NDT sensors. Five such systems have been built and deployed at various power plants. The paper deals with the development of the BARCIS system for meeting the ISI requirements of coolant channels, development cycle of this system from its conception to evolution to the present state, challenges, data generated and experience gained (ISI of nearly 900 coolant channels has been completed). Prior to BARCIS, pressure tube gauging equipment for pre-service inspection of coolant tubes was developed in 1980. Moreover a tool for ISI of coolant channels in dry condition was developed in 1990. The paper also describes evolution of various contingency procedures and devices developed over the last one decade. Future plans taking into account technological advancement, changes in the scope of inspection due to design and operating experiences and plant layout will also be covered. The paper describes the efforts put in to develop drive and control mechanism to suit the different vault layouts. The drive mechanism is responsible for linear and rotary movement of the inspection head to carry out 100% volumetric inspection. Special emphasis has been laid on the safety devices required during the inspection activity. Special measures for heavy water retention in

Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

International Nuclear Information System (INIS)

Prasser, Horst-Michael; Grunwald, Gerhard; Hoehne, Thomas; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter

2003-01-01

The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold

Evaluation of CRUDTRAN code to predict transport of corrosion products and radioactivity in the PWR primary coolant system

International Nuclear Information System (INIS)

Lee, C.B.

2002-01-01

CRUDTRAN code is to predict transport of the corrosion products and their radio-activated nuclides such as cobalt-58 and cobalt-60 in the PWR primary coolant system. In CRUDTRAN code the PWR primary circuit is divided into three principal sections such as the core, the coolant and the steam generator. The main driving force for corrosion product transport in the PWR primary coolant comes from coolant temperature change throughout the system and a subsequent change in corrosion product solubility. As the coolant temperature changes around the PWR primary circuit, saturation status of the corrosion products in the coolant also changes such that under-saturation in steam generator and super-saturation in the core. CRUDTRAN code was evaluated by comparison with the results of the in-reactor loop tests simulating the PWR primary coolant system and PWR plant data. It showed that CRUDTRAN could predict variations of cobalt-58 and cobalt-60 radioactivity with time, plant cycle and coolant chemistry in the PWR plant. (author)

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

Advances in the physics modelling of CANDU liquid injection shutdown systems

International Nuclear Information System (INIS)

Smith, H.J.; Robinson, R.; Guertin, C.

1993-01-01

The physics modelling of liquid poison injection shutdown systems in CANDU reactors accounts for the major phenomena taking place by combining the effects of both moderator hydraulics and neutronics. This paper describes the advances in the physics modelling of liquid poison injection shutdown systems (LISS), discusses some of the effects of the more realistic modelling, and briefly describes the automation methodology. Modifications to the LISS methodology have improved the realism of the physics modelling, showing that the previous methodology significantly overestimated energy deposition during the simulation of a loss of coolant transient in Bruce A, by overestimating the reactivity transient. Furthermore, the automation of the modelling process has reduced the time needed to carry put LISS evaluations to the same level as required for shutoff-rod evaluations, while at the same time minimizing the amount of input, and providing a method for tracing all files used, thus adding a level of quality assurance to the calculation. 5 refs., 11 figs

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.)

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)

Vision system for precision alignment of coolant channels

International Nuclear Information System (INIS)

Kar, S.; Rao, Y.V.; Valli Kumar; Joshi, D.G.; Chadda, V.K.; Nigam, R.K.; Kayal, J.N.; Panwar, S.; Sinha, R.K.

1997-01-01

This paper describes a vision system which has been developed for precision alignment of Coolant Channel Replacement Machine (CCRM) with respect to the front face of the coolant channel under repair/replacement. It has provisions for automatic as well as semi-automatic alignment. A special lighting scheme has been developed for providing illumination to the front face of the channel opening. This facilitates automatic segmentation of the digitized image. The segmented image is analysed to obtain the centre of the front face of the channel opening and thus the extent of misalignment i.e. offset of the camera with respect to the front face of the channel opening. The offset information is then communicated to the PLC to generate an output signal to drive the DC servo motors for precise positioning of the co-ordinate table. 2 refs., 5 figs

10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting systems. 50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND... function of these systems. High point vents are not required for the tubes in U-tube steam generators...

10 CFR 830 Major Modification Determination for Emergency Firewater Injection System Replacement

International Nuclear Information System (INIS)

Duckwitz, Noel

2011-01-01

The continued safe and reliable operation of the ATR is critical to the Department of Energy (DOE) Office of Nuclear Energy (NE) mission. While ATR is safely fulfilling current mission requirements, a variety of aging and obsolescence issues challenge ATR engineering and maintenance personnel's capability to sustain ATR over the long term. First documented in a series of independent assessments, beginning with an OA Environmental Safety and Health Assessment conducted in 2003, the issues were validated in a detailed Material Condition Assessment (MCA) conducted as a part of the ATR Life Extension Program in 2007. Accordingly, near term replacement of aging and obsolescent original ATR equipment has become important to ensure ATR capability in support of NE's long term national missions. To that end, a mission needs statement has been prepared for a non-major system acquisition which is comprised of three interdependent sub-projects. The first project will replace the existent diesel-electrical bus (E-3), switchgear, and the fifty year old antiquated marine diesels with commercial power that is backed with safety-related emergency diesel generators (EDGs), switchgear, and uninterruptible power supply. The second project will replace the four, obsolete, original primary coolant pumps and motors. The third project, the subject of this major modification determination, will replace the current emergency firewater injection system (EFIS). The replacement water injection system will function as the primary emergency water injection system with the EFIS being retained as a defense-in-depth backup. Completion of this and the two other age-related projects (replacement of the ATR diesel bus (E-3) and switchgear and replacement of the existent aged primary coolant pumps and motors) will resolve major age-related operational issues plus make a significant contribution in sustaining the ATR safety and reliability profile. The major modification criteria evaluation of the

Comparative design study of FR plants with various coolants. 1. Studies on Na coolant FR, Pb-Bi coolant FR, gas coolant FR

International Nuclear Information System (INIS)

Konomura, Mamoru; Shimakawa, Yoshio; Hori, Toru; Kawasaki, Nobuchika; Enuma, Yasuhiro; Kida, Masanori; Kasai, Shigeo; Ichimiya, Masakazu

2001-01-01

In Phase I of the Feasibility Studies on the Commercialized Fast Reactor (FR) Cycle System, plant designs on FR were performed with various coolants. This report describes the plant designs on FR with sodium, lead-bismuth, CO 2 gas and He gas coolants. A construction cost of 0.2 million yen/kWe was set up as a design goal. The result is as follows: The sodium reactor has a capability to obtain the goal, and lead-bismuth and gas reactors may satisfy the goal with further improvements. (author)

Recirculation system for nuclear reactors

International Nuclear Information System (INIS)

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

1980-01-01

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

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)

Membrane systems and their use in nuclear power plants. Treatment of primary coolant

Energy Technology Data Exchange (ETDEWEB)

Kus, Pavel; Bartova, Sarka; Skala, Martin; Vonkova, Katerina [Research Centre Rez, Husinec-Rez (Czech Republic). Technological Circuits Innovation Dept.; Zach, Vaclav; Kopa, Roman [CEZ a.s., Temelin (Czech Republic). Nuclear Power Plant Temelin

2016-03-15

In nuclear power plants, drained primary coolant containing boric acid is currently treated in the system of evaporators and by ion exchangers. Replacement of the system of evaporators by membrane system (MS) will result in lower operating cost mainly due to lower operation temperature. In membrane systems the feed primary coolant is separated into two output streams: retentate and permeate. Retentate stream consists of the concentrated boric acid solution together with other components, while permeate stream consists of purified water. Results are presented achieved by testing a pilot-plant unit of reverse osmosis in nuclear power plant (NPP) Temelin.

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

Coolant processing device for nuclear reactor

International Nuclear Information System (INIS)

Kizawa, Hideo; Funakoshi, Toshio; Izumoji, Yoshiaki

1981-01-01

Purpose: To reduce an entire facility cost by concentrating and isolating tritium accumulated in coolants, removing the tritium out of the system, and returning hydrogen gas generated at a reactor accident to a recombiner in a closed loop by the switching of a valve. Constitution: Coolant from a reactor cooling system processed by a chemical volume control system facility (CVCS) and coolant drain from various devices processed by a liquid waste disposing system facility (LWDS) are fed to a tritium isolating facility, in which they are isolated into concentrated tritium water and dilute tritium water. The concentrated tritium water is removed out of the system and stored. The dilute tritium water is reused as supply water for coolant. If an accident occurs to cause hydrogen to be generated, a closed loop is formed between the containment vessel and the recombiner, the hydrogen is recombined with oxygen in the air of the closed loop to be thus returned to water. (Kamimura, M.)

Conceptual design of primary coolant purification system using cylindrical membrane for nuclear energy system base on HTGR

International Nuclear Information System (INIS)

Piping Supriatna

2011-01-01

The recent progress of reactor technology design for next generation reactor will be implemented on cogeneration reactor, which the aim of reactor operation not only for generating electrical energy, but also for other application like desalination, industrial manufacturing process, hydrogen production, Enhanced Oil Recovery (EOR), etc. The cogeneration reactor concept developed for generate energy effectively, efficiently and sustainable, which reserve of uranium and thorium nuclear fuel for cogeneration reactor is supply able for world energy demand until next thousand years. The cogeneration reactor produce temperature output higher than commonly Nuclear Power Plant (NPP), and need special Heat Exchanger with helium gas as coolant. In order to preserve heat transfer with high efficiency, constant purity of the gas must be maintained as well as possible, especially contamination from its impurities. In this research has been designed modeling and assessment of primary coolant gas purification system with purify and fill up helium gas continuously, by using Cylindrical Helium Splitting Membrane and helium gas inventory system. The result of flow rate helium assessment for the purification system is 0.844x10 -3 kg/sec, where helium flow rate of reactor primary coolant is 120 kg/sec. The result of study show that the Primary Coolant Gas Purification System is enable to be implemented on Cogeneration Reactor HTGR200C. (author)

Numerical FEM Analyses of primary coolant system at NPP Temelin

International Nuclear Information System (INIS)

Junek, L.; Slovacek, M.; Ruzek, L.; Moulis, P.

2003-01-01

The main goal of this paper is to inform about the beginning and first steps of implementation of an aging management system at the Temelin NPP. The aging management system is important not only for achieving the current safety level but also for reaching operational reliability of a production unit equipment above the life time assumed by the original design, typically over 40 years. A method to locate the most prominent degradation regions is described. A global shell model of the primary coolant system including all loops and their components - reactor pressure vessel (RPV), steam generator (SG), main coolant pump (MCP), pressurizer, feed water and steam pipelines system is presented. The results of stress-strain analysis on the measured service parameters base are given. Validation of the results is very important and the method to compare the service measurement data with the numerical results is described. The global/local approach is mentioned and discussed. The effects of the complete global system on the individual components under monitoring are transformed into more accurate local spatial models. The local spatial models are used to analyze the gradual lifetime exhaustion of a facility during its service operation. Two spatial local models are presented, viz. feed water nozzle of SG and main coolant piping system T-brunch. The results of analysis of the local spatial models are processed by the neural network computing method, which is also described. The actual gradual damage of the material of the components under monitoring can be obtained based on the analyses performed and on the results from the neural network in combination with the knowledge of the real material characteristics. The procedures applied are included in the DIALIFE diagnostic system

Feeding and purge systems of coolant primary circuit and coolant secondary circuit control of the I sup(123) target

International Nuclear Information System (INIS)

Almeida, G.L. de.

1986-01-01

The Radiation Protection Service of IEN (Brazilian-CNEN) detected three faults in sup(123)I target cooling system during operation process for producing sup(123)I: a) non hermetic vessel containing contaminated water from primary coolant circuit; possibility of increasing radioactivity in the vessel due to accumulation of contaminators in cooling water and; situation in region used for personnels to arrange and adjust equipments in nuclear physics area, to carried out maintenance of cyclotron and target coupling in irradiation room. The primary circuit was changed by secondary circuit for target coolant circulating through coil of tank, which receive weater from secondary circuit. This solution solved the three problems simultaneously. (M.C.K.)

Multi-objective optimization of the reactor coolant system

International Nuclear Information System (INIS)

Chen Lei; Yan Changqi; Wang Jianjun

2014-01-01

Background: Weight and size are important criteria in evaluating the performance of a nuclear power plant. It is of great theoretical value and engineering significance to reduce the weight and volume of the components for a nuclear power plant by the optimization methodology. Purpose: In order to provide a new method for the optimization of nuclear power plant multi-objective, the concept of the non-dominated solution was introduced. Methods: Based on the parameters of Qinshan I nuclear power plant, the mathematical models of the reactor core, the reactor vessel, the main pipe, the pressurizer and the steam generator were built and verified. The sensitivity analyses were carried out to study the influences of the design variables on the objectives. A modified non-dominated sorting genetic algorithm was proposed and employed to optimize the weight and the volume of the reactor coolant system. Results: The results show that the component mathematical models are reliable, the modified non-dominated sorting generic algorithm is effective, and the reactor inlet temperature is the most important variable which influences the distribution of the non-dominated solutions. Conclusion: The optimization results could provide a reference to the design of such reactor coolant system. (authors)

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.)

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

The Analysis of Applying Different Coolants for Cooling Systems in the Office Building

Directory of Open Access Journals (Sweden)

Rasa Kanapienytė

2011-12-01

Full Text Available The paper analyzes air conditioning systems of different coolants on the basis of an example of a typical office building. Depending on the type of a coolant fan coil unit, active chilled beams, variable refrigerant volumes and air cooling systems were designed. The article suggests hydraulic and aerodynamic calculations and evaluates initial investments, energy expenditures and operating costs of the compared systems. Considering economic calculations, the pay-back time of the systems was assessed and the sensitivity analysis of electricity prices was carried out. The results of the conducted investigation show the most appropriate analysed system for office buildings taking into account the efficient use of electricity and initial investments.Article in Lithuanian

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)

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

Secondary coolant purification system

International Nuclear Information System (INIS)

Stiteler, F.Z.; Donohue, J.P.

1978-01-01

The present invention combines the attributes of volatile chemical addition, continuous blowdown, and full flow condensate demineralization. During normal plant operation (defined as no primary to secondary leakage) condensate from the condenser is pumped through a full flow condensate demineralizer system by the condensate pumps. Volatile chemical additions are made. Dissolved and suspended solids are removed in the condensate polishers by ion exchange and/or filtration. At the same time a continuous blowdown of approximately 1 percent of the main steaming rate of the steam generators is maintained. Radiation detectors monitor the secondary coolant. If these monitors indicate no primary to secondary leakage, the blowdown is cooled and returned directly to the condensate pump discharge. If one of the radiation monitors should indicate a primary to secondary leak, when the temperature of the effluent exiting from the blowdown heat exchanger is compatible with the resin specifications of the ion exchangers, the bypass valve causes the blowdown flow to pass through the blowdown ion exchangers

In-operation diagnostic system for reactor coolant pump

International Nuclear Information System (INIS)

Sugiyama, Mitsunobu; Hasegawa, Ichiro; Kitahara, Hiromichi; Shimamura, Kazuo; Yasuda, Chiaki; Ikeda, Yasuhiro; Kida, Yasuo.

1996-01-01

A reactor coolant pump (RCP) is one of the most important rotating machines in the primary loop nuclear power plants. To improve the reliability and of nuclear power plants, a new diagnostic system that enables early detection of RCP faults has been developed. This system is based on continuous monitoring of vibration and other process data. Vibration is an important indicator of mechanical faults providing information on physical phenomena such as changes in dynamic characteristics and excitation forces changes that signal failure or incipient failure. This new system features comparative vibration analysis and simulation to anticipate equipment failure. (author)

Characterization of primary coolant purification system samples for assay of spent ion exchanger radionuclide inventor

International Nuclear Information System (INIS)

Sajin Prasad, S.; Pant, Amar; Sharma, Ranjit; Pal, Sanjit

2018-01-01

The primary coolant system water of a research reactor contains various fission and activation products and the water is circulated continuously through ion exchange resin cartridges, to reduce the radioactive ionic impurity present in it. The coolant purification system comprises of an ion exchange cooler, two micro filters, and a battery of six ion exchanger beds, associated valves, piping and instrumentation (Heavy water System Operating manual, 2014). The spent cartridge is finally disposed off as active solid waste which contains predominantly long lived fission and activation products. The heavy water coolant is also used to cool the structural assemblies after passing through primary heat exchanger and a metallic strainer, which accumulates the fission and activation products. When there is a reduction of coolant flow through these strainers, they are removed for cleaning and decontamination. This paper describes the characterization of ion exchange resin samples and liquid effluent generated during ultra sonic decontamination of strainer. The results obtained can be used as a methodology for the assay of the spent ion exchanger cartridges radionuclide inventory, during its disposal

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.

Design of the coolant system for the Large Coil Test Facility pulse coils

International Nuclear Information System (INIS)

Bridgman, C.; Ryan, T.L.

1983-01-01

The pulse coils will be a part of the Large Coil Test Facility in Oak Ridge, Tennessee, which is designed to test six large tokamak-type superconducting coils. The pulse coil set consists of two resistive coaxial solenoid coils, mounted so that their magnetic axis is perpendicular to the toroidal field lines of the test coil. The pulse coils provide transient vertical fields at test coil locations to simulate the pulsed vertical fields present in tokamak devices. The pulse coils are designed to be pulsed for 30 s every 150 s, which results in a Joule heating of 116 kW per coil. In order to provide this capability, the pulse coil coolant system is required to deliver 6.3 L/s (100 gpm) of subcooled liquid nitrogen at 10-atm absolute pressure. The coolant system can also cool down each pulse coil from room temperature to liquid nitrogen temperature. This paper provides details of the pumping and heat exchange equipment designed for the coolant system and of the associated instrumentation and controls

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

Polonium release from an ATW burner system with liquid lead-bismuth coolant

International Nuclear Information System (INIS)

Li, N.; Yefimov, E.; Pankratov, D.

1998-04-01

The authors analyzed polonium release hazards in a conceptual pool-type ATW burner with liquid lead-bismuth eutectic (LBE) coolant. Simplified quantitative models are used based on experiments and real NPP experience. They found little Po contamination outside the burner under normal operating conditions with nominal leakage from the gas system. In sudden gas leak and/or coolant spill accidents, the P contamination level can reach above the regulation limit but short exposure would not lead to severe health consequences. They are evaluating and developing mitigation methods

Reactor coolant purification system circulation pumps (CUW pumps)

International Nuclear Information System (INIS)

Tsutsui, Toshiaki

1979-01-01

Coolant purification equipments for BWRs have been improved, and the high pressure purifying system has become the main type. The quantity of purifying treatment also changed to 2% of the flow rate of reactor feed water. As for the circulation pumps, canned motor pumps are adopted recently, and the improvements of reliability and safety are attempted. The impurities carried in by reactor feed water and the corrosion products generated in reactors and auxiliary equipments are activated by neutron irradiation or affect heat transfer adversely, adhering to fuel claddings are core structures. Therefore, a part of reactor coolant is led to the purification equipments, and returned to reactors after the impurities are eliminated perfectly. At the time of starting and stopping reactors, excess reactor water and the contaminated water from reactors are transferred to main condenser hot wells or waste treatment systems. Thus the prescribed water quality is maintained. The operational modes of and the requirements for the CUW pumps, the construction and the features of the canned motor type CUW pumps are explained. Recently, a pump operated for 11 months without any maintenance has been disassembled and inspected, but the wear of bearings has not been observed, and the high reliability of the pump has been proved. (Kako, I.)

Effect of parameter variation of reactor coolant pump on loss of coolant accident consequence

International Nuclear Information System (INIS)

Dang Gaojian; Huang Daishun; Gao Yingxian; He Xiaoqiang

2015-01-01

In this paper, the analyses were carried out on Ling'ao nuclear power station phase II to study the consequence of the loss of coolant accident when the homologous characteristic curves and free volumes of the reactor coolant pump changed. Two different pumps used in the analysis were 100D (employed on Ling'ao nuclear power station phase II) and ANDRITZ. The thermal characteristics in the large break LOCA accident were analyzed using CATHRE GB and CONPATE4, and the reactor coolant system hydraulics load during blow-clown phase of LOCA accident was analyzed using ATHIS and FORCET. The calculated results show that the homologous characteristic curves have great effect on the thermal characteristics of reactor core during the reflood phase of the large break LOCA accident. The maximum cladding surface temperatures are quite different when the pump's homologous characteristic curves change. On the other hand, the pump's free volume changing results in the variation of the LOCA rarefaction wave propagation, and therefore, the reactor coolant system hydraulic load in LOCA accident would be different. (authors)

A dynamic model of the reactor coolant system flow for KMRR plant simulation

International Nuclear Information System (INIS)

Rhee, B.W.; Noh, T.W.; Park, C.; Sim, B.S.; Oh, S.K.

1990-01-01

To support computer simulation studies for reactor control system design and performance evaluation, a dynamic model of the reactor coolant system (RCS) and reflector cooling system has been developed. This model is composed of the reactor coolant loop momentum equation, RCS pump dynamic equation, RCS pump characteristic equation, and the energy equation for the coolant inside the various components and piping. The model is versatile enough to simulate the normal steady-state conditions as well as most of the anticipated flow transients without pipe rupture. This model has been successfully implemented as the plant simulation code KMRRSIM for the Korea Multi-purpose Research Reactor and is now under extensive validation testing. The initial stage of validation has been comparison of its result with that of already validated, more detailed reactor system transient codes such as RELAP5. The results, as compared to the predictions by RELAP5 simulation, have been generally found to be very encouraging and the model is judged to be accurate enough to fulfill its intended purpose. However, this model will continue to be validated against other plant's data and eventually will be assessed by test data from KMRR

Development of lead-bismuth coolant technology for nuclear device

International Nuclear Information System (INIS)

Kamata, Kin-ya; Kitano, Teruaki; Ono, Mikinori

2004-01-01

Liquid lead-bismuth is a promising material as a future fast reactor coolant or an intensive neutron source material for accelerator driven transmutation system (ADS). To develop nuclear plants and their installations using lead-bismuth coolant for practical use, both coolant technologies, inhabitation process of steels and quality control of coolant, and total operation system for liquid lead-bismuth plants are required. Based on the experience of liquid metal coolant, Mitsui Engineering and Shipbuilding Co., Ltd. (MES) has completed the liquid lead-bismuth forced circulation loop and has acquired various engineering data on main components including economizer. As a result of tis operation, MES has developed key technologies of lead-bismuth coolant such as controlling of oxygen content in lead-bismuth and a purification of lead-bismuth coolant. MES participated in the national project, ''The Development of Accelerator Driven Transmutation System'', together with JAERI (Japan Atomic Energy Research Institute) and started corrosion test for beam window of ADS. (author)

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

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

Guidelines to achieve seals with minimal leak rates for HWR-NPR coolant system components

International Nuclear Information System (INIS)

Finn, P.A.

1991-03-01

Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs

Guidelines to achieve seals with minimal leak rates for HWR-NPR coolant system components

Energy Technology Data Exchange (ETDEWEB)

Finn, P.A.

1991-03-01

Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs.

Reactor coolant and associated systems in nuclear power plants

International Nuclear Information System (INIS)

1986-01-01

This Safety Guide outlines the design requirements for the reactor coolant and associated systems (RCAS) and the features required in order to achieve their safety functions. It covers design considerations for various reactor types and encompasses the safety aspects of the functions of the RCAS both during normal operation and following postulated initiating events, and to some extent also for decommissioning

First Study of Helium Gas Purification System as Primary Coolant of Co-Generation Reactor

International Nuclear Information System (INIS)

Piping Supriatna

2009-01-01

The technological progress of NPP Generation-I on 1950’s, Generation-II, Generation-III recently on going, and Generation-IV which will be implemented on next year 2025, concept of nuclear power technology implementation not only for generate electrical energy, but also for other application which called cogeneration reactor. Commonly the type of this reactor is High Temperature Reactor (HTR), which have other capabilities like Hydrogen production, desalination, Enhanced Oil Recovery (EOR), etc. The cogeneration reactor (HTR) produce thermal output higher than commonly Nuclear Power Plant, and need special Heat Exchanger with helium gas as coolant. In order to preserve heat transfer with high efficiency, constant purity of the gas must be maintained as well as possible, especially contamination from its impurities. In this report has been done study for design concept of HTR primary coolant gas purification system, including methodology by sampling He gas from Primary Coolant and purification by using Physical Helium Splitting Membrane. The examination has been designed in physical simulator by using heater as reactor core. The result of study show that the of Primary Coolant Gas Purification System is enable to be implemented on cogeneration reactor. (author)

Li-ion battery thermal runaway suppression system using microchannel coolers and refrigerant injections

Science.gov (United States)

Bandhauer, Todd M.; Farmer, Joseph C.

2016-11-08

A battery management system with thermally integrated fire suppression includes a multiplicity of individual battery cells in a housing; a multiplicity of cooling passages in the housing within or between the multiplicity of individual battery cells; a multiplicity of sensors operably connected to the individual battery cells, the sensors adapted to detect a thermal runaway event related to one or more of the multiplicity of individual battery cells; and a management system adapted to inject coolant into at least one of the multiplicity of cooling passages upon the detection of the thermal runaway event by the any one of the multiplicity of sensors, so that the thermal runaway event is rapidly quenched.

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

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

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

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

Modular 3-D solid finite element model for fatigue analyses of a PWR coolant system

International Nuclear Information System (INIS)

Garrido, Oriol Costa; Cizelj, Leon; Simonovski, Igor

2012-01-01

Highlights: ► A 3-D model of a reactor coolant system for fatigue usage assessment. ► The performed simulations are a heat transfer and stress analyses. ► The main results are the expected ranges of fatigue loadings. - Abstract: The extension of operational licenses of second generation pressurized water reactor (PWR) nuclear power plants depends to a large extent on the analyses of fatigue usage of the reactor coolant pressure boundary. The reliable estimation of the fatigue usage requires detailed thermal and stress analyses of the affected components. Analyses, based upon the in-service transient loads should be compared to the loads analyzed at the design stage. The thermal and stress transients can be efficiently analyzed using the finite element method. This requires that a 3-D solid model of a given system is discretized with finite elements (FE). The FE mesh density is crucial for both the accuracy and the cost of the analysis. The main goal of the paper is to propose a set of computational tools which assist a user in a deployment of modular spatial FE model of main components of a typical reactor coolant system, e.g., pipes, pressure vessels and pumps. The modularity ensures that the components can be analyzed individually or in a system. Also, individual components can be meshed with different mesh densities, as required by the specifics of the particular transient studied. For optimal accuracy, all components are meshed with hexahedral elements with quadratic interpolation. The performance of the model is demonstrated with simulations performed with a complete two-loop PWR coolant system (RCS). Heat transfer analysis and stress analysis for a complete loading and unloading cycle of the RCS are performed. The main results include expected ranges of fatigue loading for the pipe lines and coolant pump components under the given conditions.

Calorimetric and reactor coolant system flow uncertainty

International Nuclear Information System (INIS)

Bates, L.; McLean, T.

1991-01-01

This paper describes a methodology for the quantification of errors associated with the determination of a feedwater flow, secondary power, and Reactor Coolant System (RCS) flow used at the Trojan Nuclear Plant to ensure compliance with regulatory requirements. The sources of error in Plant indications and process measurement are identified and tracked, using examples, through the mathematical processes necessary to calculate the uncertainty in the RCS flow measurement. An error of approximately 1.4 percent is calculated for secondary power. This error results, along with the consideration of other errors, in an uncertainty of approximately 3 percent in the RCS flow determination

Standardized sampling system for reactor coolants

International Nuclear Information System (INIS)

Divine, J.R.; Munson, L.F.; Nelson, J.L.; McDowell, R.L.; Jankowski, M.W.

1982-09-01

A three-pronged approach was developed to reach the objectives of acceptable coolant sampling, assessment of occupational exposure from corrosion products, and model development for the transport and buildup of corrosion products. Emphasis is on sampler design

Analysis of an AP600 intermediate-size loss-of-coolant accident

Energy Technology Data Exchange (ETDEWEB)

Boyack, B.E.; Lime, J.F. [Los Alamos National Lab., NM (United States)

1995-09-01

A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

Reactor auxiliary cooling facility and coolant supplying method therefor

International Nuclear Information System (INIS)

Ando, Koji; Kinoshita, Shoichiro.

1996-01-01

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

Reactor auxiliary cooling facility and coolant supplying method therefor

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1996-06-07

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

Study on 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)

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

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

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.)

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.)

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.

On-Line Coolant Chemistry Analysis

International Nuclear Information System (INIS)

LM Bachman

2006-01-01

Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level

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.)

Water quality control device and water quality control method for reactor primary coolant system

International Nuclear Information System (INIS)

Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

1995-01-01

The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

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

International Nuclear Information System (INIS)

Kim, Yun Il; Woo, Tae Ho

2017-01-01

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

Phenomena occurring in the reactor coolant system during severe core damage accidents

International Nuclear Information System (INIS)

Malinauskas, A.P.

1989-01-01

The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. In the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. In contrast to normal operating conditions, severe core damage accidents are characterized by significant temporal and spatial variations in heat and mass fluxes, and by eventual geometrical changes within the RCS. Furthermore, the difficulties in describing the system in the severe accident mode are compounded by the occurrence of chemical reactions. These reactions can influence both the thermal and the mass transport behavior of the system. In addition, behavior of the reactor vessel internals and of materials released from the core region (especially the radioactive fission products) in the course of the accident likewise become of concern to the analyst. This report addresses these concerns. 9 refs., 1 tab

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

Lead coolant test facility systems design, thermal hydraulic analysis and cost estimate

Energy Technology Data Exchange (ETDEWEB)

Khericha, Soli, E-mail: slk2@inel.gov [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Harvego, Edwin; Svoboda, John; Evans, Robert [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Dalling, Ryan [ExxonMobil Gas and Power Marketing, Houston, TX 77069 (United States)

2012-01-15

The Idaho National Laboratory prepared a preliminary technical and functional requirements (T and FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed below: Bullet Develop and demonstrate feasibility of submerged heat exchanger. Bullet Develop and demonstrate open-lattice flow in electrically heated core. Bullet Develop and demonstrate chemistry control. Bullet Demonstrate safe operation. Bullet Provision for future testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimated. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 Degree-Sign C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2008-01-01

This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. 1.2. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1981), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1986), which are superseded by this new Safety Guide. 1.3. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1981 and 1986, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2000, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included

Triboengineering problems of lead coolant in innovative fast reactors

International Nuclear Information System (INIS)

Beznosov, A.V.; Novozhilova, O.O.; Shumilkov, A.I.; Lvov, A.V.; Bokova, T.A.; Makhov, K.A.

2013-01-01

Graphical abstract: Models of experimental sites for research of processes tribology in heavy liquid metal coolant. -- Highlights: • The contact a pair of heavy liquid metal coolant for reactors on fast neutrons. • The hydrostatic bearings main circulation pumps. • Oxide coating and degree of wear of friction surfaces in heavy liquid metal coolant. -- Abstract: So far, there are plenty of works dedicated to studying the phenomenon of friction. However, there are none dedicated to functioning of contact pairs in heavy liquid-metal coolants for fast neutron, reactor installations (Kogaev and Drozdov, 1991; Modern Tribology, 2008; Drozdov et al., 1986). At the Nizhny Novgorod State Technical University, such research is conducted in respect to friction, bearings of main circulating pumps, interaction of sheaths of neutron absorber rods with their covers, of the reactor control and safety system, refueling systems, and interaction of coolant flows with, channel borders. As a result of experimental studies, the characteristic of friction pairs in the heavy, liquid metal coolant shows the presence dependences of oxide film on structural materials of the wear. The inapplicability of existing calculation methods for assessing the performance of the bearing nodes, in the heavy liquid metal coolant is shown

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

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.)

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

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

Design and fabrication of magnetic coolant filter

Science.gov (United States)

Prashanth, B. N.

2017-07-01

Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

Reactor coolant pump monitoring and diagnostic system

International Nuclear Information System (INIS)

Singer, R.M.; Gross, K.C.; Walsh, M.; Humenik, K.E.

1990-01-01

In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs

ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

Directory of Open Access Journals (Sweden)

Veijo KAUPPINEN

2012-11-01

Full Text Available Strained environment is a global problem. In metal industries the use of coolant has become more problematic in terms of both employee health and environmental pollution. It is said that the use of coolant forms approximately 8 - 16 % of the total production costs.The traditional methods that use coolants are now obviously becoming obsolete. Hence, it is clear that using a dry cutting system has great implications for resource preservation and waste reduction. For this purpose, a new cooling system is designed for dry cutting. This paper presents the new eco-friendly cooling innovation and the benefits gained by using this method. The new cooling system relies on a unit for ionising ejected air. In order to compare the performance of using this system, cutting experiments were carried out. A series of tests were performed on a horizontal turning machine and on a horizontal machining centre.

Main coolant pump testing at Ontario Hydro

International Nuclear Information System (INIS)

Hartlen, R.

1991-01-01

This article describes Ontario Hydro Research Division's experience with a computerized data acquisition and analysis system for monitoring mechanical vibration in reactor coolant pumps. The topics covered include bench-marking of the computer system and the coolant pumps, signatures of normal and malfunctioning pumps, analysis of data collected by the monitoring system, simulation of faults, and concerns that have been expressed about data interpretation, sensor types and locations, alarm/shutdown limits and confirmation of nondestructive examination testing. This presentation consists of overheads only

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)

SQL injection detection system

OpenAIRE

Vargonas, Vytautas

2017-01-01

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

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.

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)

Staged fuel and air injection in combustion systems of gas turbines

Energy Technology Data Exchange (ETDEWEB)

Hughes, Michael John; Berry, Jonathan Dwight

2018-04-10

A gas turbine including a working fluid flowpath extending aftward from a forward injector in a combustor. The combustor may include an inner radial wall, an outer radial wall, and, therebetween, a flow annulus, and a third radial wall formed about the outer radial wall that forms an outer flow annulus. A staged injector may intersect the flow annulus so to attain an injection point within the working fluid flowpath by which aftward and forward annulus sections are defined. Air directing structure may include an aftward intake section corresponding to the aftward annulus section and a forward intake section corresponding to the forward annulus section. The air directing structure may include a switchback coolant flowpath to direct air from the compressor discharge cavity to the staged injector. The switchback coolant flowpath may include an upstream section through the flow annulus, and a downstream section through the outer flow annulus.

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)

Recent developments in coolant systems for Indus Accelerator Complex at RRCAT, Indore

International Nuclear Information System (INIS)

Nanda, Dipankar; Tiwari, Bablu; Pandey, R.M.

2015-01-01

Scarcity of fresh water forces mankind to explore other possible water sources that can meet the increasing demand of coolants in industries, R and D sectors and other establishments where water is used as coolant. It also becomes a challenge for water chemist to control water chemistry to keep the equipments/devices intact during its operation using water as coolant. Deionised (DI) and soft water have been used as coolants for Indus Accelerator Complex, RRCAT, Indore. DI water is produced and its quality is maintained either by conventional ion exchange method or a hybrid method of membrane separation and ion exchange technique. This requires handling of corrosive chemicals, manpower, space for plant installation, and a long array of water treatment units. CSL has implemented the idea of rain water harvesting to produce DI water after systematic studies in laboratory. The concerning issues are reduced to almost one-fourth by using rain water to produce DI water. The harvesting system has been in use for last three years. Heat is dissipated into air by evaporation of soft water in cooling tower. Requirement of soft water makeup has been estimated to be about 40,000 ltrs. / day (max.) if the machine is operated at its designed specifications. Non-availability of soft water (which circulates in open loop) may lead to shut down like situation and looking for alternate source becomes quite essential. Laboratory studies (water analysis and treatment) on sewage water (available 1,00,000 ltrs/day) from RRCAT colony as a possible source of producing soft water show promising result. (author)

Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor coolant system and systems decontamination. Summary status report. Volume 1

International Nuclear Information System (INIS)

Doerge, D.H.; Miller, R.L.; Scotti, K.S.

1986-05-01

This document summarizes information relating to the decontamination and restoration of the Three Mile Island Unit 2 reactor coolant system and other plant systems. Data have been collected from activity reports, reactor containment entry records, and other sources and entered in a computerized data system which permits extraction/manipulation of specific data which can be used in planning for recovery from a loss of coolant event similar to that experienced by the Three Mile Island Unit 2 on March 28, 1979. This report contains a summary of radiation exposures, manpower, and time spent in radiation areas during the referenced period. Support activities conducted outside of radiation areas are not included. Computer reports included are: A chronological listing of all activities related to decomtamination and restoration of the reactor coolant system and other plant systems for the period of April 5, 1979, through December 19, 1984; a summary of labor and exposures by department for the same time period; and summary reports for each major task undertaken in connection with this specific work scope during the referenced time period

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

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

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

HANARO secondary coolant management

International Nuclear Information System (INIS)

Kim, Seon Duk.

1998-02-01

In this report, the basic theory for management of water quality, environmental factors influencing to the coolant, chemicals and its usage for quality control of coolant are mentioned, and water balance including the loss rate by evaporation (34.3 m 3 /hr), discharge rate (12.665 m 3 /hr), concentration ratio and feed rate (54.1 m 3 /hr) are calculated at 20 MW operation. Also, the analysis data of HANSU Limited for HANARO secondary coolant (feed water and circulating coolant) - turbidity, pH, conductivity, M-alkalinity, Ca-hardness, chloride ion, total iron ion, phosphoric ion and conversion rate are reviewed. It is confirmed that the feed water has good quality and the circulating coolant has been maintained within the control specification in general, but some items exceeded the control specification occasionally. Therefore it is judged that more regular discharge of coolant is needed. (author). 6 refs., 17 tabs., 18 figs

Evaluation of alternate secondary (and tertiary) coolants for the molten-salt breeder reactor

International Nuclear Information System (INIS)

Kelmers, A.D.; Baes, C.F.; Bettis, E.S.; Brynestad, J.; Cantor, S.; Engel, J.R.; Grimes, W.R.; McCoy, H.E.; Meyer, A.S.

1976-04-01

The three most promising coolant selections for an MSBR have been identified and evaluated in detail from the many coolants considered for application either as a secondary coolant in 1000-MW(e) MSBR configurations using only one coolant, or as secondary and tertiary coolants in an MSBR dual coolant configuration employing two different coolants. These are, as single secondary coolants: (1) a ternary sodium--lithium--beryllium fluoride melt; (2) the sodium fluoroborate--sodium fluoride eutectic melt, the present reference design secondary coolant. In the case of the dual coolant configuration, the preferred system is molten lithium--beryllium fluoride (Li 2 BeF 4 ) as the secondary coolant and helium gas as the tertiary coolant

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

Fuel-Coolant Interactions: Visualization and Mixing Measurements

International Nuclear Information System (INIS)

Loewen, Eric P.; Bonazza, Riccardo; Corradini, Michael L.; Johannesen, Robert E.

2002-01-01

Dynamic X-ray imaging of fuel-coolant interactions (FCI), including quantitative measurement of fuel-coolant volume fractions and length scales, has been accomplished with a novel imaging system at the Nuclear Safety Research Center at the University of Wisconsin, Madison. The imaging system consists of visible-light high-speed digital video, low-energy X-ray digital imaging, and high-energy X-ray digital imaging subsystems. The data provide information concerning the melt jet velocity, melt jet configuration, melt volume fractions, void fractions, and spatial and temporal quantification of premixing length scales for a model fuel-coolant system of molten lead poured into a water pool (fuel temperatures 500 to 1000 K; jet diameters 10 to 30 mm; coolant temperatures 20 to 90 deg. C). Overall results indicate that the FCI has three general regions of behavior, with the high fuel-coolant temperature region similar to what might be expected under severe accident conditions. It was observed that the melt jet leading edge has the highest void fraction and readily fragments into discrete masses, which then subsequently subdivide into smaller masses of length scales <10 mm. The intact jet penetrates <3 to 5 jet length/jet diameter before this breakup occurs into discrete masses, which continue to subdivide. Hydrodynamic instabilities can be visually identified at the leading edge and along the jet column with an interfacial region that consists of melt, vapor, and water. This interface region was observed to grow in size as the water pool temperature was increased, indicating mixing enhancement by boiling processes

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

Mathematical model of the reactor coolant pump

International Nuclear Information System (INIS)

Kozuh, M.

1989-01-01

The mathematical model of reactor coolant pump is described in this paper. It is based on correlations for centrifugal reactor coolant pumps. This code is one of the elements needed for the simulation of the whole NPP primary system. In subroutine developed according to this model we tried in every possible detail to incorporate plant specific data for Krsko NPP. (author)

Analysis of the VVER-1000 coolant transient benchmark phase 1 with the code system RELAP5/PARCS

International Nuclear Information System (INIS)

Victor Hugo Sanchez Espinoza

2005-01-01

Full text of publication follows: As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during

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

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)

Design of coolant distribution system (CDS) for ITER PF AC/DC converter

Energy Technology Data Exchange (ETDEWEB)

Guo, Bin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Zhiquan, E-mail: zhquansong@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Fu, Peng; Xu, Xuesong; Li, Chuan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wang, Min; Dong, Lin [China International Nuclear Fusion Energy Program Execution Center, Beijing 100862 (China)

2016-10-15

Highlights: • System process and arrangement has been proposed to meet the multiple requirements from the converter system. • Thermal hydraulic analysis model has been developed to size and predict the system operation behavior. • Prototype test has been performed to validate the proposed design methodology. - Abstract: The Poloidal Field (PF) converter unit, playing an essential role in the plasma shape and position control in vertical and horizontal direction, which is an important part of ITER power supply system. As an important subsystem of the converter unit, the coolant distribution system has the function to distribute the cooling water from ITER component cooling water system (CCWS) to its main components at the required flow rate, pressure and temperature. This paper presents the thermal hydraulic design of coolant distribution system for the ITER PF converter unit. Different operational requirements of the PF converter unit regarding flow rate, temperature and pressure have been analyzed to design the system process and arrangement. A thermal-hydraulic analysis model has been built to size the system and predict the flow rate and temperature distribution of the system under the normal operation. Based on the system thermal-hydraulic analysis results, the system pressure profile has been plotted to evaluate the pressure behavior along each client flow path. A CDS prototype for the ITER PF converter has been constructed and some experiments have been performed on it. A good agreement of the flow distribution and temperature behavior between the simulated and test results validate the proposed design methodology.

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

RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

International Nuclear Information System (INIS)

Kawai, H.

2001-01-01

JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

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

Performance investigation of an automotive car radiator operated with nanofluid-based coolants (nanofluid as a coolant in a radiator)

International Nuclear Information System (INIS)

Leong, K.Y.; Saidur, R.; Kazi, S.N.; Mamun, A.H.

2010-01-01

Water and ethylene glycol as conventional coolants have been widely used in an automotive car radiator for many years. These heat transfer fluids offer low thermal conductivity. With the advancement of nanotechnology, the new generation of heat transfer fluids called, 'nanofluids' have been developed and researchers found that these fluids offer higher thermal conductivity compared to that of conventional coolants. This study focused on the application of ethylene glycol based copper nanofluids in an automotive cooling system. Relevant input data, nanofluid properties and empirical correlations were obtained from literatures to investigate the heat transfer enhancement of an automotive car radiator operated with nanofluid-based coolants. It was observed that, overall heat transfer coefficient and heat transfer rate in engine cooling system increased with the usage of nanofluids (with ethylene glycol the basefluid) compared to ethylene glycol (i.e. basefluid) alone. It is observed that, about 3.8% of heat transfer enhancement could be achieved with the addition of 2% copper particles in a basefluid at the Reynolds number of 6000 and 5000 for air and coolant respectively. In addition, the reduction of air frontal area was estimated.

Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide (Spanish Edition)

International Nuclear Information System (INIS)

2010-01-01

This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1982), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1987), which are superseded by this new Safety Guide. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1982 and 1987, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2004, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included.

Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

International Nuclear Information System (INIS)

Hunsbedt, A.; Boardman, C.E.

1993-01-01

A dual passive cooling system for liquid metal cooled nuclear fission reactors is described, comprising the combination of: a reactor vessel for containing a pool of liquid metal coolant with a core of heat generating fissionable fuel substantially submerged therein, a side wall of the reactor vessel forming an innermost first partition; a containment vessel substantially surrounding the reactor vessel in spaced apart relation having a side wall forming a second partition; a first baffle cylinder substantially encircling the containment vessel in spaced apart relation having an encircling wall forming a third partition; a guard vessel substantially surrounding the containment vessel and first baffle cylinder in spaced apart relation having a side wall forming a forth partition; a sliding seal at the top of the guard vessel edge to isolate the dual cooling system air streams; a second baffle cylinder substantially encircling the guard vessel in spaced part relationship having an encircling wan forming a fifth partition; a concrete silo substantially surrounding the guard vessel and the second baffle cylinder in spaced apart relation providing a sixth partition; a first fluid coolant circulating flow course open to the ambient atmosphere for circulating air coolant comprising at lent one down comer duct having an opening to the atmosphere in an upper area thereof and making fluid communication with the space between the guard vessel and the first baffle cylinder and at least one riser duct having an opening to the atmosphere in the upper area thereof and making fluid communication with the space between the first baffle cylinder and the containment vessel whereby cooling fluid air can flow from the atmosphere down through the down comer duct and space between the forth and third partitions and up through the space between the third and second partition and the riser duct then out into the atmosphere; and a second fluid coolant circulating flow

Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde

International Nuclear Information System (INIS)

Cardenas V, J.; Mugica R, C. A.; Godinez S, V.

2013-10-01

In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft 2 (4.6 cm 2 ), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

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

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

Reference is made to coolant channels for pressurised water and boiling water reactors and the arrangement described aims to improve heat transfer between the fuel rods and the coolant. Baffle means extending axially within the channel are provided and disposed relative to the fuel rods so as to restrict flow oscillations occurring within the coolant from being propagated transversely to the axis of the channel. (UK)

Reactor primary coolant system pipe rupture study. Progress report No. 33, January--June 1975

International Nuclear Information System (INIS)

1975-10-01

The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase 1), analytical and experimental efforts (Phase 2) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue crack growth rate studies focused on LWR primary piping materials in a simulated BWR primary coolant environment, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, (c) studies directed at quantifying weld sensitization in Type 304 stainless steel, (d) support studies to characterize the electrochemical potential behavior of a typical BWR primary water environment and (e) special tests related to simulation of fracture surfaces characteristic of IGSCC field failures

Residual heat removal pump and low pressure safety injection pump retrofit program

International Nuclear Information System (INIS)

Dudiak, J.G.; McKenna, J.M.

1992-01-01

Residual Heat Removal (RHR) and low pressure safety injection (LPSI) pumps installed in pressurized water-to-reactor power plants are used to provide low-head safety injection in the event of loss of coolant in the reactor coolant system. Because these pumps are subjected to rather severe temperature and pressure transients, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. Typically the pump impeller is mounted on an extended motor shaft (close-coupled configuration) and a mechanical seal is employed at the pump end of the shaft. Traditionally RHR and LPSI pumps have been a significant maintenance item for many utilities. Periodic mechanical seal of motor bearing replacement often is considered routine maintenance. The closed-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. This paper introduces a design modification developed to convert the close-coupled RHR and LPSI pumps to a coupled configuration

Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

International Nuclear Information System (INIS)

Sanchez-Espinoza, Victor Hugo

2008-07-01

As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

Energy Technology Data Exchange (ETDEWEB)

Sanchez-Espinoza, Victor Hugo

2008-07-15

As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

Lamp system with conditioned water coolant and diffuse reflector of polytetrafluorethylene(PTFE)

Science.gov (United States)

Zapata, Luis E.; Hackel, Lloyd

1999-01-01

A lamp system with a very soft high-intensity output is provided over a large area by water cooling a long-arc lamp inside a diffuse reflector of polytetrafluorethylene (PTFE) and titanium dioxide (TiO.sub.2) white pigment. The water is kept clean and pure by a one micron particulate filter and an activated charcoal/ultraviolet irradiation system that circulates and de-ionizes and biologically sterilizes the coolant water at all times, even when the long-arc lamp is off.

Integrated Mechanical Pulse Jet Coolant Delivery System Performance for Minimal Quantity Lubrication

OpenAIRE

Nik Fazli Sapian; Badrul Omar; Mohd Hamdi Abd Shukor

2010-01-01

Minimum quantity lubrication (MQL) machining is one of the promising solutions to the requirement for decrease in cutting fluid consumption. This research describes MQL machining in a range of lubricant consumption 2.0ml/h, which is 10–100 times smaller than the consumption usually adopted in industries. MQL machining in this range is called pulse jet coolant delivery system in this research. A specially designed system was used for concentrating small amounts of lubricant onto the cutting in...

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.)

Evaluation of primary coolant leaks and assessment of detection methods

International Nuclear Information System (INIS)

Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

1986-01-01

A review of the French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all actual leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by compliance with the criteria defined in the operating technical specifications

Evaluation of primary coolant leaks and assessment of detection methods

International Nuclear Information System (INIS)

Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

1984-11-01

A review of French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all occurred leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by the compliance with the criteria defined in the operating technical specifications

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

Science.gov (United States)

Permana, Sidik

2017-07-01

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

Low pressure injection sequence sensitivity study of the M1 module of MEDICI

International Nuclear Information System (INIS)

Corradini, M.L.; Moses, G.A.; Norkus, J.K.; Welzbacker, R.T.

1985-01-01

In order to assess the consequences of a PWR containment failure and the ensuing radiological source term following a severe reactor accident, it is necessary to understand the ex-vessel behavior of the molten core. The M1 module of MEDICI models the dynamic fuel-coolant mixing, energetic interaction, and ejection of fuel and coolant from the reactor cavity following such an accident. A sensitivity study of the low pressure injection sequence was performed utilizing a Box-Behnken statistical design to treat five sets of input variables considered to be significant in the mixing and steam explosion processes. The low pressure injection sequence was studied in which the molten corium is modeled as a pour stream entering the cavity without entraining or sweeping out fuel or coolant

Single failure effects of reactor coolant system large bore hydraulic snubbers for Korean Standard Nuclear Power Plant

International Nuclear Information System (INIS)

Choi, T.S.; Park, S.H.; Sung, K.K.; Kim, T.W.; Jheon, J.H.

1996-01-01

A potential snubber single failure is one of the safety significances identified in General Safety Issue 113 for Large Bore Hydraulic Snubber (LBHS) dynamic qualification. This paper investigates dynamic structural effects of single failures of the steam generator and reactor coolant pump snubbers in Korean Standard Nuclear Power Plant by performing the time history dynamic analyses for the reactor coolant system under seismic and postulated pipe break events. The seismic input motions considered are the synthesized ground time histories conforming to SRP 3.7.1, and he postulated pipe break input loadings result from steam generator main seam line and feedwater line pipe breaks which govern pipe breaks remaining after applying LBB to the main coolant line and primary side ranch lines equal to and greater than 12 inch nominal pipe size

Primary system hydraulic characteristics after modification of reactor coolant pumps' impeller wheels at Bohunice NPP executed in 2012 and 2013

International Nuclear Information System (INIS)

Hermansky, Jozef; Zavodsky, Martin

2014-01-01

A coolant flow through the reactor is usually determined after annual outages at Slovak NPP (VVER 440) in two distinct ways. First method is determination on the basis of the secondary system parameters - measurement of thermal balances. The value achieved by this method is used as the input parameter in the Table of allowed reactor operation modes. The second method draws from the primary system parameters - measurement of primary system hydraulic characteristics. Flow nozzles used for the measurement of feed water flow behind high pressure heaters were replaced at both Bohunice Units during outages in 2008. The feed water flow behind high pressure heaters is one of the main parameters used for the determination of coolant flow through the reactor by the first method. Compared to the measurement executed during previous fuel cycles, the calculated coolant flow through the reactor decreased considerably after the change of flow nozzles. The imaginary change of coolant flow through the reactor at Unit 3 was -1,6 %; and at Unit 4 -2,6 %. This change was not proved by the parallel measurement of primary system hydraulic characteristics. Later it was found out that the original flow nozzles used for 25 years were substantially deposited (original inner diameter of the nozzles was reduced by about 0,6-0,9 mm). Therefore feed water flow measurement was untrustworthy within the recent years. On the findings stated above, Bohunice NPP has decided to increase coolant flow through the reactor by changing the reactor coolant pumps impeller wheels. The modification of impellers wheels is planned within years 2012 to 2014. During the outages in 2013 two impeller wheels were replaced at both units. Nowadays Unit 4 is operated with all 6 new impeller wheels and Unit 3 with four new impeller wheels. Modification of last two impeller wheels at Unit 3 will be performed during the outage in 2014. On account of impeller wheels modification, non-standard measurement of PS hydraulic

IEA-R1 renewed primary coolant piping system stress analysis

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

2015-01-01

A partial replacement of the IEA-R1 piping system was conducted in 2014. The aim of this work is to perform the stress analysis of the renewed primary piping system of the IEA-R1, taking into account the as built conditions and the pipe modifications. The nuclear research reactor IEA-R1 is a pool type reactor designed by Babcox-Willcox, which is operated by IPEN since 1957. The primary coolant system is responsible for removing the residual heat of the Reactor core. As a part of the life management, a regular inspection detected some degradation in the primary piping system. In consequence, part of the piping system was replaced. The partial renewing of the primary piping system did not imply in major piping layout modifications. However, the stress condition of the piping systems had to be reanalyzed. The structural stress analysis of the primary piping systems is now presented and the final results are discussed. (author)

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)

Theoretical Study on the Flow of Refilling Stage in a Safety Injection Tank

Energy Technology Data Exchange (ETDEWEB)

Park, Jun Sang [Halla Univ. Daejeon (Korea, Republic of)

2017-10-15

In this study, a theoretical analysis was performed to the flow of refilling stage in a safety injection tank, which is the core cooling system of nuclear power plant in an emergency. A theoretical model was proposed with a nonlinear governing equation defining on the flow of the refilling process of the coolant. Utilizing the Taylor-series expansion, the 1st - order approximation flow equation was obtained, along with its analytic solution of closed type, which could predict accurately the variations of free surface height and flow rate of the coolant. The availability of theoretical result was confirmed by comparing with previous experimental results.

THE RHIC INJECTION SYSTEM.

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-29

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

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.)

ITER Neutral Beam Injection System

International Nuclear Information System (INIS)

Ohara, Yoshihiro; Tanaka, Shigeru; Akiba, Masato

1991-03-01

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

AGING MANAGMENT OF REACTOR COOLANT SYSTEM MECHANICAL COMPONENTS FOR LICENSE RENEWAL

International Nuclear Information System (INIS)

SUBUDHI, M.; MORANTE, R.; LEE, A.D.

2002-01-01

The reactor coolant system (RCS) mechanical components that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer. steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions. determination of the effects of aging on their intended safety functions. and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. In addition, this paper discusses time-limited aging analyses associated with neutron embrittlement of the reactor vessel beltline region and thermal fatigue

Multi-state reliability for coolant pump based on dependent competitive failure model

International Nuclear Information System (INIS)

Shang Yanlong; Cai Qi; Zhao Xinwen; Chen Ling

2013-01-01

By taking into account the effect of degradation due to internal vibration and external shocks. and based on service environment and degradation mechanism of nuclear power plant coolant pump, a multi-state reliability model of coolant pump was proposed for the system that involves competitive failure process between shocks and degradation. Using this model, degradation state probability and system reliability were obtained under the consideration of internal vibration and external shocks for the degraded coolant pump. It provided an effective method to reliability analysis for coolant pump in nuclear power plant based on operating environment. The results can provide a decision making basis for design changing and maintenance optimization. (authors)

Laser-based sensor for a coolant leak detection in a nuclear reactor

Science.gov (United States)

Kim, T.-S.; Park, H.; Ko, K.; Lim, G.; Cha, Y.-H.; Han, J.; Jeong, D.-Y.

2010-08-01

Currently, the nuclear industry needs strongly a reliable detection system to continuously monitor a coolant leak during a normal operation of reactors for the ensurance of nuclear safety. In this work, we propose a new device for the coolant leak detection based on tunable diode laser spectroscopy (TDLS) by using a compact diode laser. For the feasibility experiment, we established an experimental setup consisted of a near-IR diode laser with a wavelength of about 1392 nm, a home-made multi-pass cell and a sample injection system. The feasibility test was performed for the detection of the heavy water (D2O) leaks which can happen in a pressurized heavy water reactor (PWHR). As a result, the device based on the TDLS is shown to be operated successfully in detecting a HDO molecule, which is generated from the leaked heavy water by an isotope exchange reaction between D2O and H2O. Additionally, it is suggested that the performance of the new device, such as sensitivity and stability, can be improved by adapting a cavity enhanced absorption spectroscopy and a compact DFB diode laser. We presume that this laser-based leak detector has several advantages over the conventional techniques currently employed in the nuclear power plant, such as radiation monitoring, humidity monitoring and FT-IR spectroscopy.

Improvement of Measurement Accuracy of Coolant Flow in a Test Loop

Energy Technology Data Exchange (ETDEWEB)

Hong, Jintae; Kim, Jong-Bum; Joung, Chang-Young; Ahn, Sung-Ho; Heo, Sung-Ho; Jang, Seoyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In this study, to improve the measurement accuracy of coolant flow in a coolant flow simulator, elimination of external noise are enhanced by adding ground pattern in the control panel and earth around signal cables. In addition, a heating unit is added to strengthen the fluctuation signal by heating the coolant because the source of signals are heat energy. Experimental results using the improved system shows good agreement with the reference flow rate. The measurement error is reduced dramatically compared with the previous measurement accuracy and it will help to analyze the performance of nuclear fuels. For further works, out of pile test will be carried out by fabricating a test rig mockup and inspect the feasibility of the developed system. To verify the performance of a newly developed nuclear fuel, irradiation test needs to be carried out in the research reactor and measure the irradiation behavior such as fuel temperature, fission gas release, neutron dose, coolant temperature, and coolant flow rate. In particular, the heat generation rate of nuclear fuels can be measured indirectly by measuring temperature variation of coolant which passes by the fuel rod and its flow rate. However, it is very difficult to measure the flow rate of coolant at the fuel rod owing to the narrow gap between components of the test rig. In nuclear fields, noise analysis using thermocouples in the test rig has been applied to measure the flow velocity of coolant which circulates through the test loop.

Evaluation of a postulated loss of coolant accident (LOCA) due to a 160 cm2 break in a cold leg of Angra 2 nuclear power plant

International Nuclear Information System (INIS)

Azevedo, Carlos Vicente Goulart de; Palmieri, Elcio Tadeu; Aronne, Ivan Dionysio

2002-01-01

The development of a qualified full nodalization of Angra2 NPP for RELAP5/Mod 3.2.2 gamma, aiming at the evaluation of a comprehensive number of accidents and transients, thus providing suitable safety analysis support for licensing purposes, is being carried out within the framework of CNEN internal technical cooperation, involving some of its institutes (CDTN, IPEN and IEN) and the Reactors Coordination (CODRE). This work presents a simulation of a postulated Angra2 small cold leg break loss of coolant accident (SBLOCA). A 160 cm 2 break is supposed to occur at one cold leg between the main coolant pump and the reactor vessel and is described in the Angra2 Final Safety Analysis Report, section 15.6.4.1.3.4. The simulation of several types of transients and accidents is necessary to verify the adequate performance of the modeled logic and systems. In general, the analysis of such and accident allows to demonstrate the safety Injection System performance and the reliable transition between the high pressure safety injection, the accumulator injection and the residual heat removal phases. Furthermore, it is assumed that some components are out of service due to fail or repair in order to make a conservative analysis. The results showed a compatible behavior of the molded systems and that the simulated Emergency Core Cooling System was able to provide sufficient cooling to avoid any damage to the core. (author)

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

Reactor coolant system and containment aqueous chemistry

International Nuclear Information System (INIS)

Torgerson, D.F.

1986-01-01

Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

Decontamination of primary cooling system

International Nuclear Information System (INIS)

Morikawa, Yoshitake.

1985-01-01

Purpose: To effectively eliminate radioactivity accumulated in pipeways, equipments, etc in primary coolant circuits of BWR type power plants by utilizing ion displacement reactions. Method: The reactor pressure vessel is connected with a feedwater pipeway, steam pipeway and a recycling pipeway. The recycling pipeway is disposed with a recycling pump. A recycling by-pass line is branched from the recycling pipeway and disposed with a recycling system heat exchanger and chemical injection point. Water is filled in the primary coolant and heated 280 0 C. Then, while maintaining water at that temperature, non-radioactive cobalt ions are injected and circulated within the system, by which radioactivity accumulated in pipeways, equipments or the likes can effectively be removed. (Horiuchi, T.)

Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

Energy Technology Data Exchange (ETDEWEB)

Yu, Seon Oh; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Sung Joong [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

2017-08-15

The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

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)

Preliminary Performance Analysis Program Development for Safety System with Safeguard Vessel

International Nuclear Information System (INIS)

Kang, Han-Ok; Lee, Jun; Park, Cheon-Tae; Yoon, Ju-Hyeon; Park, Keun-Bae

2007-01-01

SMART is an advanced modular integral type pressurized water reactor for a seawater desalination and an electricity production. Major components of the reactor coolant system such as the pressurizer, Reactor Coolant Pump (RCP), and steam generators are located inside the reactor vessel. The SMART can fundamentally eliminate the possibility of large break loss of coolant accidents (LBLOCAs), improve the natural circulation capability, and better accommodate and thus enhance a resistance to a wide range of transients and accidents. The safety goals of the SMART are enhanced through highly reliable safety systems such as the passive residual heat removal system (PRHRS) and the safeguard vessel coupled with the passive safety injection feature. The safeguard vessel is a steel-made, leak-tight pressure vessel housing the RPV, SIT, and the associated valves and pipelines. A primary function of the safeguard vessel is to confine any radioactive release from the primary circuit within the vessel under DBAs related to loss of the integrity of the primary system. A preliminary performance analysis program for a safety system using the safeguard vessel is developed in this study. The developed program is composed of several subroutines for the reactor coolant system, passive safety injection system, safeguard vessel including the pressure suppression pool, and PRHRS. A small break loss of coolant accident at the upper part of a reactor is analyzed and the results are discussed

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.)

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

Coolant voiding analysis following SGTR for an HLMC reactor

International Nuclear Information System (INIS)

Farmer, M.T.; Spencer, B.W.; Sienicki, J.J.

2000-01-01

Concepts are under development at Argonne National Laboratory for a small, modular, proliferation-resistant nuclear power steam supply system. Of primary interest here is the simplified system design, featuring steam generators that are directly immersed in the lead-bismuth eutectic (LBE) coolant of the primary system. To support the safety case for this design approach, model development and analysis of transient coolant voiding during a postulated guillotine-type steam generator tube rupture event has been carried out. For the current design, the blowdown will occur from the steam generator shell into the ruptured 12.7-mm-inside-diameter tube through which the LBE coolant passes. The steam will expand biaxially in the tube, with a portion of the flow vented upward to eventually expand into the cover-gas region, while the balance of the flow is vented downward as a jet into the surrounding downward-flowing LBE. Coolant freezing is not an issue in this case because of high feedwater temperature in relation to the freezing point of the LBE. The specific objectives of the current work are to (a) determine the penetration behavior of the steam jet into the lower cold-leg region, (b) characterize the resultant void behavior in terms of coherent bubble versus breakup into a size distribution of small bubbles, and (c) characterize the motion of the bubbles with regard to rise to the cover-gas region (via the liner-to-coolant vessel gap) versus downward transport with the flowing LBE and subsequent upflow through the core to the cover-gas region

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.)

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

International Nuclear Information System (INIS)

Han, S. J.; Im, H. K.; Yang, J. U.

2003-01-01

Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). To use RIA, the present study focuses on the detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study is to evaluate the success criteria of Aggressive Secondary Cooldown (ASC) in Small Break Loss Of Coolant Accident (SBLOCA) with total loss of High Pressure Safety Injection (HPSI) and to enhance the understanding of related thermal hydraulic behavior and phenomena. The accident scenario was 2 inch coldleg break LOCA without HPSI, with 1/2 Low Pressure Safety Injection (LPSI), and performing ASC limited by 55.6 .deg. C /hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip, which successively reaches the LPSI condition for about 1.5hr after starting ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria 1204.4 .deg. C (2200 .deg. F). In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that operator should maintain the adequate ASC operation. However, it is necessary to evaluate uncertainties arisen from the related parameters of the ASC operation

The installation welding of pressure water reactor coolant piping

International Nuclear Information System (INIS)

Deng Feng

2010-01-01

Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

Sloshing of coolant in a seismically isolated reactor

International Nuclear Information System (INIS)

Wu, T.S.; Guildys, J.; Seidensticker, R.W.

1988-01-01

During a seismic event, the liquid coolant inside the reactor vessel has sloshing motion which is a low-frequency phenomenon. In a reactor system incorporated with seismic isolation, the isolation frequency usually is also very low. There is concern on the potential amplification of sloshing motion of the liquid coolant. This study investigates the effects of seismic isolation on the sloshing of liquid coolant inside the reactor vessel of a liquid metal cooled reactor. Based on a synthetic ground motion whose response spectra envelop those specified by the NRC Regulator Guide 1.60, it is found that the maximum sloshing wave height increases from 18 in. to almost 30 in. when the system is seismically isolated. Since higher sloshing wave may introduce severe impact forces and thermal shocks to the reactor closure and other components within the reactor vessel, adequate design considerations should be made either to suppress the wave height or to reduce the effects caused by high waves

Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

International Nuclear Information System (INIS)

Duce, S.W.; Marley, A.W.; Freeman, A.L.

1989-12-01

Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

HTGR-GT primary coolant transient resulting from postulated turbine deblading

International Nuclear Information System (INIS)

Cadwallader, G.J.; Deremer, R.K.

1980-11-01

The turbomachine is located within the primary coolant system of a nuclear closed cycle gas turbine plant (HTGR-GT). The deblading of the turbine can cause a rapid pressure equilibration transient that generates significant loads on other components in the system. Prediction of and design for this transient are important aspects of assuring the safety of the HTGR-GT. This paper describes the adaptation and use of the RATSAM program to analyze the rapid fluid transient throughout the primary coolant system during a spectrum of turbine deblading events. Included are discussions of (1) specific modifications and improvements to the basic RATSAM program, which is also briefly described; (2) typical results showing the expansion wave moving upstream from the debladed turbine through the primary coolant system; and (3) the effect on the transient results of different plenum volumes, flow resistances, times to deblade, and geometries that can choke the flow

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

Directory of Open Access Journals (Sweden)

A.S. Ekariansyah

2016-08-01

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

Organic coolant for ARIES-III

International Nuclear Information System (INIS)

Sze, D.K.; Sviatoslavsky, I.; Sawan, M.; Gierszewski, P.; Hollies, R.; Sharafat, S.; Herring, S.

1991-04-01

ARIES-III is a D-He 3 reactor design study. It is found that the organic coolant is well suited for the D-He 3 reactor. This paper discusses the unique features of the D-He 3 reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs

Primary coolant system of BWR type reactor

International Nuclear Information System (INIS)

Ibe, Hidefumi; Takahashi, Masanori; Aoki, Yasuko

1997-01-01

The present invention provides a water quality control system for preventing corrosion and for extending working life of structural materials of a BWR-type reactor. Namely, a sensor group 1 and a sensor group 2 are disposed at different positions such as in a feedwater system, a recycling system, main steam pipes, and a pressure vessel, respectively. Each sensor group can record and generate alarms independently. The sensor group 1 for usual monitoring is connected to a calculation device by way of a switch to confirm that the monitored values are within a proper range by the injection of a water quality moderating agent. The sensor group 2 is caused to stand alone or connected with the calculation device by way of a switch optionally. When abnormality should occur in the sensor group 1, the sensor group 2 determines the limit for the increase/decrease of controlling amount of the moderating agent at a portion where the conditions are changed to the most severe direction by using data base. The moderating agent is injected and controlled based on the controlling amount. The system of the present invention can optionally cope with a new sensor and determination for new water quality standards. Then the evaluation/control accuracy of the entire system can be improved while covering up the errors of each sensor. (I.S.)

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)

The sodium coolant

International Nuclear Information System (INIS)

Rodriguez, G.

2004-01-01

The sodium is the best appropriate coolant for the fast neutrons reactors technology. Thus the fast neutrons reactors development is intimately bound to the sodium technology. This document presents the sodium as a coolant point of view: atomic structure and characteristics, sodium impacts on the fast neutron reactors technology, chemical properties of the sodium and the consequences, quality control in a nuclear reactor, sodium treatment. (A.L.B.)

Loss-of-coolant accident analysis of the Savannah River new production reactor design

International Nuclear Information System (INIS)

Maloney, K.J.; Pryor, R.J.

1990-11-01

This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. 5 refs., 50 figs., 4 tabs

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.

Design technology development of the main coolant pump for an integral reactor

International Nuclear Information System (INIS)

Park, J. S.; Lee, J. S.; Kim, M. H.; Kim, D. W.; Kim, J. I.

2004-01-01

All of the reactor coolant pump currently used in commercial nuclear power plant were imported from foreign country. Now, the developing program of design technology for the reactor coolant pump will be started in a few future by domestic researchers. At this stage, the design technology of the main coolant pump for an integral reactor is developed based on the regulation of domestic nuclear power plant facilities. The main coolant pump is a canned motor axial pump, which accommodates all constraints required from the integral reactor system. The main coolant pump does not have mechanical seal device because the rotor of motor and the shaft of impeller are the same one. There is no flywheel on the rotating shaft of main coolant pump so that the coastdown duration time is short when the electricity supply is cut off

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)

On-line real time gamma analysis of primary coolant

International Nuclear Information System (INIS)

Kalechstein, W.; Kupca, S.; Lipsett, J.J.

1985-10-01

The evolution of failed fuel monitoring at CANDU power stations is briefly summarized and the design of the latest system for failed fuel detection at a multi-unit power station is described. At each reactor, the system employs a germanium spectrometer combined with a novel spectrum analyzer that simultaneously accumulates the gamma-ray spectrum of the coolant and provides the control room with the concentration of radioisotope activity in the coolant for the gaseous fission products Xe-133, Xe-135, Kr-88 and I-131 in real time and with statistical precision independent of count rate. A gross gamma monitor is included to provide independent information on the level of radioactivity in the coolant and extend the measurement range at very high count rates. A central computer system archives spectra received from all four spectrum analyzers and provides both the activity concentrations and the release rates of specified isotopes. Compared with previous systems the current design offers improvements in that the activity concentrations are updated much more frequently, improved tools are provided for long term surveillance of the heat transport system and the monitor is more reliable and less costly

Sodium coolant of fast reactors: Experience and problems

International Nuclear Information System (INIS)

Kozlov, F.A.; Volchkov, L.G.; Drobyshev, A.V.; Nikulin, M.P.; Kochetkov, L.A.; Alexeev, V.V.

1997-01-01

In present report the following subjects are considered: state of the coolant and sodium systems under normal operating condition as well as under decommissioning, disclosing of sodium circuits and liquidation of its consequences, cleaning from sodium and decontamination under repairing works of equipment and circuits. Cleaning of coolant and sodium systems under normal operating conditions and under accident contamination. Cleaning of the equipment under repairing works and during decommissioning from sodium and products of its interaction with water and air. Treatment of sodium waste, taking into account a possibility of sodium fires. It is shown that the state of coolant, cover gas, surfaces of constructive materials which are in contact with them, cleaning systems, formed during installation operation require development of specific technologies. Developed technologies ensured safety operation of sodium cooled installations as in normal operating conditions so in abnormal situations. R and D activities in this field and experience gained provided a solid base for coping with problems arising during decommissioning. Prospective research problems are emphasized where the future efforts should be concentrated in order to improve characteristics of sodium cooled reactors and to make their decommissioning optimal and safe. (author)

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

International Nuclear Information System (INIS)

Yeon-Jong Yoo; Qiao Wu; James J Sienicki

2005-01-01

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

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

Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

Directory of Open Access Journals (Sweden)

Seon Oh Yu

2017-08-01

Full Text Available The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

Dual fuel injection piggyback controller system

Science.gov (United States)

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

2017-09-01

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

Minimizing secondary coolant blowdown in HANARO

International Nuclear Information System (INIS)

Park, Y. C.; Woo, J. S.; Ryu, J. S.; Cho, Y. G.; Lim, N. Y.

2000-01-01

There is about 80m 3 /h loss of the secondary cooling water by evaporation, windage and blowdown during the operation of HANARO, 30MW research reactor. The evaporation and the windage is necessary loss to maintain the performance of cooling tower, but the blowdown is artificial lose to get rid of the foreign material and to maintain the quality of the secondary cooling water. Therefore, minimizing the blowdown loss was studied. It was confirmed, through the relation of the number of cycle and the loss rate of secondary coolant, that the number of cycle is saturated to 12 without blowdown because of the windage loss. When the secondary coolant is treated by high Ca-hardness treatment program (the number of cycle > 10) to maintain the number of cycle around 12 without blowdown, only the turbidity exceeds the limit. By adding filtering system it was confirmed, through the relation of turbidity and filtering rate of secondary cooling water, that the turbidity is reduced below the limit (5 deg.) by 2% of filtering rate without blowdown. And it was verified, through the performance test of back-flow filtering unit, that this unit gets rid of foreign material up to 95% of the back-flow and that the water can be reused as coolant. Therefore, the secondary cooling water can be treated by the high Ca-hardness program and filter system without blowdown

Statistical analysis of the Ft. Calhoun reactor coolant pump system

International Nuclear Information System (INIS)

Patel, Bimal; Heising, C.D.

1997-01-01

In engineering science, statistical quality control techniques have traditionally been applied to control manufacturing processes. An application to commercial nuclear power plant maintenance and control is presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) of the Ft. Calhoun nuclear power plant. This research uses capability analysis, Shewhart X-bar, R charts, canonical correlation methods, and design of experiments to analyze the process for the state of statistical control. The results obtained show that six out of ten parameters are under control specification limits and four parameters are not in the state of statistical control. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with ample time to respond to possible emergency situations and thus improve plant safety and reliability. (Author)

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)

International Space Station Active Thermal Control Sub-System On-Orbit Pump Performance and Reliability Using Liquid Ammonia as a Coolant

Science.gov (United States)

Morton, Richard D.; Jurick, Matthew; Roman, Ruben; Adamson, Gary; Bui, Chinh T.; Laliberte, Yvon J.

2011-01-01

The International Space Station (ISS) contains two Active Thermal Control Sub-systems (ATCS) that function by using a liquid ammonia cooling system collecting waste heat and rejecting it using radiators. These subsystems consist of a number of heat exchangers, cold plates, radiators, the Pump and Flow Control Subassembly (PFCS), and the Pump Module (PM), all of which are Orbital Replaceable Units (ORU's). The PFCS provides the motive force to circulate the ammonia coolant in the Photovoltaic Thermal Control Subsystem (PVTCS) and has been in operation since December, 2000. The Pump Module (PM) circulates liquid ammonia coolant within the External Active Thermal Control Subsystem (EATCS) cooling the ISS internal coolant (water) loops collecting waste heat and rejecting it through the ISS radiators. These PM loops have been in operation since December, 2006. This paper will discuss the original reliability analysis approach of the PFCS and Pump Module, comparing them against the current operational performance data for the ISS External Thermal Control Loops.

Iron crud supply device to reactor coolant

International Nuclear Information System (INIS)

Baba, Takao.

1993-01-01

In a device for supplying iron cruds into reactor coolants in a BWR type power plant, a system in which feed water containing iron cruds is supplied to the reactor coolants after once passing through an ion exchange resin is disposed. As a result, iron cruds having characteristics similar with those of naturally occurring iron cruds in the plant are obtained and they react with ionic radioactivity, to form composite oxides. Then, iron cruds having high performance of being secured to the surface of a fuel cladding tube can be supplied to the reactor coolants, thereby enabling to greatly reduce the density of reactor water ionic radioactivity. In its turn, dose rate on the surface of pipelines can be reduced, thereby enabling to reduce operators' radiation exposure dose in the plant. Further, contamination of a condensate desalting device due to iron cruds can be prevented, and further, the density of the iron cruds supplied can easily be controlled. (N.H.)

Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

Energy Technology Data Exchange (ETDEWEB)

Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

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)

Radiation leakage monitoring method and device from primary to secondary coolant systems in nuclear reactor

International Nuclear Information System (INIS)

Tajiri, Yoshiaki; Umehara, Toshihiro; Yamada, Masataka.

1993-01-01

The present invention monitors radiation leaked from any one of primary cooling systems to secondary cooling systems in a plurality of steam generators. That is, radiation monitoring means each corresponding to steam each generators are disposed to the upstream of a position where main steam pipes are joined. With such a constitution, since the detection object of each of radiation monitoring means is secondary coolants before mixing with secondary coolants of other secondary loops or dilution, lowering of detection accuracy can be avoided. Except for the abnormal case, that is, a case neither of radiation leakage nor of background change, the device is adapted as a convenient measuring system only with calculation performance. Once abnormality occurs, a loop having a value exceeding a standard value is identified by a single channel analyzer function. The amount of radiation leakage from the steam generator belonging to the specified loop is monitored quantitatively by a multichannel analyzer function. According to the method of the present invention, since specific spectrum analysis is conducted upon occurrence of abnormality, presence of radiation leakage and the scale thereof can be judged rapidly. (I.S.)

Formation and hydraulic effects of deposits in high temperature sodium coolant systems

International Nuclear Information System (INIS)

Yunker, W.

1976-01-01

Deposition of sodium impurities in the high temperature (600 0 C), high flow (Reynolds Number approximately equal to 8 x 10 4 ) regions of a sodium coolant circuit is being studied to determine its possible hydraulic effects. Increases in flow impedance (pressure drop/volume flow 2 ) of up to 30 percent have been detected in an annular flow sensor. The apparatus and preliminary results of these tests are presented. Continuing tests are to specifically identify the materials involved and the system conditions under which the formations occur

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

International Nuclear Information System (INIS)

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

2001-01-01

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

The development of robotic system for inspecting and repairing NPP primary coolant system of high-level radioactive environment

Energy Technology Data Exchange (ETDEWEB)

Kim, Seung Ho; Kim, Ki Ho; Jung, Seung Ho; Kim, Byung Soo; Hwang, Suk Yeoung; Kim, Chang Hoi; Seo, Yong Chil; Lee, Young Kwang; Lee, Yong Bum; Cho, Jai Wan; Lee, Jae Kyung; Lee, Yong Deok

1997-07-01

This project aims at developing a robotic system to automatically handle inspection and maintenance of NPP safety-related facilities in high-level radioactive environment. This robotic system under development comprises two robots depending on application fields - a mobile robot and multi-functional robot. The mobile robot is designed to be used in the area of primary coolant system during the operation of NPP. This robot enables to overcome obstacles and perform specified tasks in unstructured environment. The multi-functional robot is designed for performing inspection and maintenance tasks of steam generator and nuclear reactor vessel during the overhaul periods of NPP. Nuclear facilities can be inspected and repaired all the time by use of both the mobile robot and the multi-functional robot. Human operator, by teleoperation, monitors the movements of such robots located at remote task environment via video cameras and controls those remotely generating desired commands via master manipulator. We summarize the technology relating to the application of the mobile robot to primary coolant system environment, the applicability of the mobile robot through 3D graphic simulation, the design of the mobile robot, the design of its radiation-hardened controller. We also describe the mechanical design, modeling, and control system of the multi-functional robot. Finally, we present the design of the force-reflecting master and the modeling of virtual task environment for a training simulator. (author). 47 refs., 16 tabs., 43 figs.

The development of robotic system for inspecting and repairing NPP primary coolant system of high-level radioactive environment

International Nuclear Information System (INIS)

Kim, Seung Ho; Kim, Ki Ho; Jung, Seung Ho; Kim, Byung Soo; Hwang, Suk Yeoung; Kim, Chang Hoi; Seo, Yong Chil; Lee, Young Kwang; Lee, Yong Bum; Cho, Jai Wan; Lee, Jae Kyung; Lee, Yong Deok.

1997-07-01

This project aims at developing a robotic system to automatically handle inspection and maintenance of NPP safety-related facilities in high-level radioactive environment. This robotic system under development comprises two robots depending on application fields - a mobile robot and multi-functional robot. The mobile robot is designed to be used in the area of primary coolant system during the operation of NPP. This robot enables to overcome obstacles and perform specified tasks in unstructured environment. The multi-functional robot is designed for performing inspection and maintenance tasks of steam generator and nuclear reactor vessel during the overhaul periods of NPP. Nuclear facilities can be inspected and repaired all the time by use of both the mobile robot and the multi-functional robot. Human operator, by teleoperation, monitors the movements of such robots located at remote task environment via video cameras and controls those remotely generating desired commands via master manipulator. We summarize the technology relating to the application of the mobile robot to primary coolant system environment, the applicability of the mobile robot through 3D graphic simulation, the design of the mobile robot, the design of its radiation-hardened controller. We also describe the mechanical design, modeling, and control system of the multi-functional robot. Finally, we present the design of the force-reflecting master and the modeling of virtual task environment for a training simulator. (author). 47 refs., 16 tabs., 43 figs

Comparative analysis of coolants for FBR of future nuclear power

International Nuclear Information System (INIS)

Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I.

2001-01-01

Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR

Comparative analysis of coolants for FBR of future nuclear power

Energy Technology Data Exchange (ETDEWEB)

Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I. [State Scientific Center of Russian Federation, Institute for Physics and Power Engineering named after Academician A.I. Leipusky, Kaluga Region (Russian Federation)

2001-07-01

Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR.

Design criteria of primary coolant chemistry in SMART-P

International Nuclear Information System (INIS)

Choi, Byung Seon; Kim, Ah Young; Kim, Seong Hoon; Yoon, Ju Hyeon; Zee, Sung Qunn

2005-01-01

SMART-P differs significantly from commercially designed PWRs. Materials inventories used in SMART-P differ from that at PWRs. All surfaces of the primary circuit with the primary coolant are either made from or plated with stainless steel. The material of steam generator (SG) is also different from that of the standard material of the commercially operating PWRs: titanium alloy for the steam generator tubes. Also, SMART-P primary coolant technology differs from that in PWRs: ammonia is used as a pH raising agent and hydrogen formed due to radiolytic processes is kept in specific range by ammonia dosing. Nevertheless, main objectives of the SMART-P primary coolant are the same as at PWRs: to assure primary system pressure boundary integrity, fuel cladding integrity and to minimize out-of-core radiation buildup. The objective of this work is to introduce the design criteria for the primary water chemistry for SMART-P from the viewpoint of the system characteristics and the chemical design concept

A volumetric flow sensor for automotive injection systems

International Nuclear Information System (INIS)

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

2008-01-01

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

A volumetric flow sensor for automotive injection systems

Science.gov (United States)

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

2008-04-01

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

Influence of Zn injection on corrosion behavior and oxide film characteristics of 304 stainless steel in borated and lithiated high temperature water

International Nuclear Information System (INIS)

Wu, Xinqiang; Liu, Xiahe; Han, En-Hou; Ke, Wei

2012-09-01

Water chemistry of the reactor coolant system plays a major role in maintaining safety and reliability of light water reactor nuclear power plants (NPPs). Zn water chemistry into pressurized water reactors (PWRs) in order to reduce the radiation buildup in primary coolant system has been widely applied, and the reduction effect has been experimentally confirmed. Zn injection can also lessen the corrosion phenomena in high temperature pressurized water by changing oxide films formed on components materials. Both the radiation buildup and material corrosion resistance in PWR coolant system are closely dependent on the oxide films formed. However, the influence of Zn injection on the chemical composition and structure of the oxide films on their protective properties is still a matter of considerable debate. The influence of Zn injection on corrosion inhibition and environmental degradation has not been fully clarified yet. Therefore, the understanding of corrosion behaviour, oxide film characteristics and their protective property is of significance to clarify the environmentally assisted material failure problems in NPPs. In the present work, oxide films formed on nuclear-grade 304 SS exposed to borated and lithiated high temperature water environments at 300 deg. C up to 4000 h with or without 10 ppb Zn injection were investigated ex-situ. Without Zn injection, the oxide films mainly consisted of Fe 3 O 4 and FeCr 2 O 4 . With Zn injection, ZnFe 2 O 4 and ZnCr 2 O 4 were detected in the oxide films at the initial stage of immersion and ZnCr 2 O 4 became dominant after long-term immersion. It was believed that the above Zn-Fe and Zn-Cr spinel oxides were formed by substitution reactions between Zn 2+ and Fe 2+ . At the initial stage of immersion, water chemistry significantly affected the formation of the oxide films. Once a stable oxide film formed, it is rather difficult to change its structure through changing water chemistry. The potential-pH diagrams for Zn

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)

Seismic analysis of the reactor coolant system of PWR nuclear power plants

International Nuclear Information System (INIS)

Borsoi, L.; Sollogoub, P.

1986-01-01

For safety considerations, seismic analyses are performed of the Reactor Coolant System (R.C.S.) of PWR Plants. After a brief description of the R.C.S. and R.C.S. operation, the paper presents the two types of analysis used to determine the effect of earthquake on the R.C.S.: modal spectral analysis and nonlinear time history analysis. The paper finally shows how seismic loadings are combined with other types of loadings and illustrates how the consideration of seismic loads affects R.C.S. design [fr

CONCEPTUAL DESIGN OF THE NSLS-II INJECTION SYSTEM.

Energy Technology Data Exchange (ETDEWEB)

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

2007-06-25

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

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)

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)

Transient behaviour of main coolant pump in nuclear power plants

International Nuclear Information System (INIS)

Delja, A.

1986-01-01

A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)

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)

Emergency core cooling systems

International Nuclear Information System (INIS)

Kubokoya, Takashi; Okataku, Yasukuni.

1984-01-01

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

Refurbishment of the IEAR1 primary coolant system piping supports

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

2015-01-01

A partial replacement of the IEA-R1 piping system was concluded in 2014. This paper presents the study and the structural analysis of the IEA-R1 primary circuit piping supports, considering all the changes involved in the replacement. The IEA-R1 is a nuclear reactor for research purposes designed by Babcox-Willcox that is operated by IPEN since 1957. The reactor life management and modernization program is being conducted for the last two decades and already resulted in a series of changes, especially on the reactor coolant system. This set of components, divided in primary and secondary circuit, is responsible for the circulation of water into the core to remove heat. In the ageing management program that includes regular inspection, some degradation was observed in the primary piping system. As result, the renewing of the piping system was conducted in 2014. Moreover the poor condition of some original piping supports gave rise to the refurbishment of all piping supports. The aim of the present work is to review the design of the primary system piping supports taking into account the current conditions after the changes and refurbishment. (author)

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

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

Examination of a failed reactor coolant pump rotating assembly from Crystal River Unit 3

International Nuclear Information System (INIS)

Hayner, G.O.; Lubnow, T.; Clary, M.

1990-01-01

On January 18, 1989, the A reactor coolant pump rotating assembly at the Crystal River Unit 3 Nuclear Power Plant failed during operation. A rotating assembly from this pump had previously failed in 1986. The reactor coolant pump was fabricated by Byron Jackson Pump Division of Borg-Warner Ind. Products, Inc. from UNS S66286 superalloy (Alloy A286). A root cause failure analysis examination was performed on the pump shaft and other components. The failure analysis included shaft vibrational mode and stress analyses, pump clearance and alignment analyses, and detailed destructive examination of the shaft and hydrostatic bearing assemblies. Based on the detailed physical examination of the shaft it was concluded that cracks initiated in the pump shaft at two sites approximately 180 0 apart in a band of shallow, thermally induced fatigue cracks. The cracks initiated at the bottom edge of the motor end shrink fit pad under the shrink fit sleeve supporting the hydrostatic bearing journal. The band of thermally induced fatigue cracks was apparently caused by mixing of cold seal injection water and hot reactor coolant in gaps between the pump shaft and sleeve. The motor end shrink fit was apparently not effective in preventing introduction of the seal injection water to this area. Initial crack propagation occurred by fatigue due to lateral vibration; however, the majority of crack propagation occurred by abnormal torsional fatigue loading induced by contact and sticking between the rotating and stationary portions of the hydrostatic bearing. Final fracture of the shaft occurred by torsional overload. Metallurgical characteristics and mechanical properties of the shaft were within design specification and probably did not significantly influence the cracking process

Addition of soluble and insoluble neutron absorbers to the reactor coolant system of TMI-2

International Nuclear Information System (INIS)

Hansen, R.F.; Silverman, J.; Queen, S.P.; Ryan, R.F.; Austin, W.E.

1984-07-01

The physical and chemical properties of six elements were studied and combined with cost estimates to determine the feasibility of adding them to the TMI-2 reactor coolant to depress k/sub eff/ to less than or equal to 0.95. Both soluble and insoluble forms of the elements B, Cd, Gd, Li, Sm, and Eu were examined. Criticality calculations were performed by Oak Ridge National Laboratory to determine the absorber concentration required to meet the 0.95 k/sub eff/ criterion. The conclusion reached is that all elements with the exception of boron have overriding disadvantages which preclude their use in this reactor. Solubility experiments in the reactor coolant show that boron solubility is the same as that of boron in pure aqueous solutions of sodium hydroxide and boric acid; consequently, solubility is not a limiting factor in reaching the k/sub eff/ criterion. Examination of the effect of pH on sodium requirements and costs for processing to remove radionuclides revealed a sharp dependence; small decreases in pH lead to a large decrease in both sodium requirements and processing costs. Boron addition to meet any contemplated reactor safety requirements can be accomplished with existing equipment; however, this addition must be made with the reactor coolant system filled and pressurized to ensure uniform boron concentration

Injection and Dump Systems

CERN Document Server

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

2012-01-01

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

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

Cleaning of aluminum after machining with coolants

International Nuclear Information System (INIS)

Roop, B.

1992-01-01

An x-ray photoemission spectroscopic study was undertaken to compare the cleaning of the Advanced Photon Source (APS) aluminum extrusion storage ring vacuum chambers after machining with and without water soluble coolants. While there was significant contamination left by the coolants, the cleaning process was capable of removing the residue. The variation of the surface and near surface composition of samples machined either dry or with coolants was negligible after cleaning. The use of such coolants in the machining process is therefore recommended

Statistical analysis of the Ft. Calhoun reactor coolant pump system

International Nuclear Information System (INIS)

Heising, Carolyn D.

1998-01-01

In engineering science, statistical quality control techniques have traditionally been applied to control manufacturing processes. An application to commercial nuclear power plant maintenance and control is presented that can greatly improve plant safety. As a demonstration of such an approach to plant maintenance and control, a specific system is analyzed: the reactor coolant pumps (RCPs) of the Ft. Calhoun nuclear power plant. This research uses capability analysis, Shewhart X-bar, R-charts, canonical correlation methods, and design of experiments to analyze the process for the state of statistical control. The results obtained show that six out of ten parameters are under control specifications limits and four parameters are not in the state of statistical control. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators Such a system would provide operators with ample time to respond to possible emergency situations and thus improve plant safety and reliability. (author)

Identification of limiting case between DBA and SBDBA (CL break area sensitivity): A new model for the boron injection system

Energy Technology Data Exchange (ETDEWEB)

Gonzalez Gonzalez, R., E-mail: r.gonzalez@ing.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122 San Piero a Grado, Pisa (Italy); Petruzzi, A., E-mail: a.petruzzi@ing.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122 San Piero a Grado, Pisa (Italy); D’Auria, F., E-mail: f.dauria@ing.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122 San Piero a Grado, Pisa (Italy); Mazzantini, O., E-mail: mazzantini@na-sa.com.ar [Nucleo-electrica Argentina Sociedad Anonima (NA-SA), Buenos Aires (Argentina)

2014-08-15

Atucha-2 is a Siemens-designed Pressurized Heavy Water Reactor (PHWR) reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarity (e.g. oblique Control Rods, Positive Void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank. The boron clouds reconstruction is obtained using a Computational Fluid Dynamics (CFD) CFX code calculation. A complete Large Break Loss Of Coolant Accident (LBLOCA) calculation implies the application of the RELAP5-3D{sup ©} system code. Within the framework of the third Agreement “Nucleoelèctrica Argentina-Sociedad Anonima (NA-SA) – University of Pisa/GRNSPG” (Contract, 2009), a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions.

Report on the Survey of the Design Review of New Reactor Applications. Volume 4: Reactor Coolant and Associated Systems

International Nuclear Information System (INIS)

Downey, Steven; Monninger, John; Nevalainen, Janne; Joyer, Philippe; Koley, Jaharlal; Kawamura, Tomonori; Chung, Yeon-Ki; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Monninger, John; Choi, Young-Joon; )

2017-01-01

At the tenth meeting of the Committee on Nuclear Regulatory Activities (CNRA) Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the licensing process survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This report provides a discussion of the survey responses related to the Reactor Coolant and Associated Systems category. The Reactor Coolant and Associated Systems category includes the following technical topics: overpressure protection, reactor coolant pressure boundary, reactor vessel, and design of the reactor coolant system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - It is common to consider operating experience and lessons learnt from the current fleet during the review process. - The most commonly and consistently identified technical expertise needed to perform design reviews related to this category are mechanical engineering and materials engineering. The complete survey

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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.)

Improvements of primary coolant shutdown chemistry and reactor coolant system cleanup

International Nuclear Information System (INIS)

Gaudard, G.; Gilles, B.; Mesnage, F.; Cattant, F.

2002-01-01

In the framework of a radiation exposure management program entitled >, EDF aims at decreasing the mass dosimetry of nuclear power plants workers. So, the annual dose per unit, which has improved from 2.44 m.Sv in 1991 to 1.08 in 2000, should target 0.8 mSv in the year 2005 term in order to meet the results of the best nuclear operators. One of the guidelines for irradiation source term reduction is the optimization of operation parameters, including reactor coolant system (RCS) chemistry in operation, RCS shutdown chemistry and RCS cleanup improvement. This paper presents the EDF strategy for the shutdown and start up RCS chemistry optimization. All the shutdown modes have been reviewed and for each of them, the chemical specifications will be fine tuned. A survey of some US PWRs shutdown practices has been conducted for an acid and reducing shutdown chemistry implementation test at one EDF unit. This survey shows that deviating from the EPRI recommended practice for acid and reducing shutdown chemistry is possible and that critical path impact can be minimized. The paper also presents some investigations about soluble and insoluble species behavior and characterization; the study focuses here on 110m Ag, 122 Sb, 124 Sb and iodine contamination. Concerning RCS cleanup improvement, the paper presents two studies. The first one highlights some limited design modifications that are either underway or planned, for an increased flow rate during the most critical periods of the shutdown. The second one focuses on the strategy EDF envisions for filters and resins selection criteria. Matching the study on contaminants behavior with the study of filters and resins selection criteria should allow improving the cleanup efficiency. (authors)

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

New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

DEFF Research Database (Denmark)

Kolaei, Alireza Rezania; Rosendahl, Lasse

2012-01-01

the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The threedimensional governing equations for the fluid flow and the heat......The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find...... heat sink configurations reduces the coolant pumping power in the system....

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.)

Evaluation of primary coolant pH operation methods for the domestic PWRs

International Nuclear Information System (INIS)

Paek, Seung Woo; Na, Jung Won; Kim, Yong Eak; Bae, Jae Heum

1992-01-01

Radioactive nuclides deposited on out-of-core surface after the radiation in the core by the transport of corrosion products (CRUD) through the primary coolant system in PWR which is the major plant type in Korea, are leading sources of radiation exposure to plant maintenance personnel. Thus, the optimal chemistry operation method is required for the reduction of radiation exposure by the corrosion products. This study analysed the actual water chemistry operation data of four operating domestic PWRs. And in order to evaluate the coolant chemistry operation data, a computer code which can calculate the activity buildup in the various chemistry conditions of PWR coolant was employed. Through the analysis of comparison between the activity buildup of actual water chemistry operation mode and that of assumed Elevated Li operation mode calculated by the computer code, it was found that the out-of-core radioactivity can be reduced by diminishing the deposition of corrosion products on the core in case that the Elevated Li operation mode is applied to the coolant chemistry operation of PWR. And the higher coolant pH operation was shown to have the advantage of the reduction of out-of-core activity buildup if the integrity of system structural materials and fuel cladding is guaranteed. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-07-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

An investigation of fluid mixing with safety injection in advanced reactors

International Nuclear Information System (INIS)

Cha, Jong Hee; Won, Soon Yean; Chung, Moon Ki; Jun, Hyung Gil

1994-01-01

The objective of this work is to investigate the fluid mixing phenomena in aspect of pressurized thermal shock(PTS) in an advanced PWR vessel downcomer during transient cooldown with safety injection. It provides comparison of fluid mixing characteristics between AP 600 DVI, designed by Westinghouse, and ABB CE System 80+ DVI, and the effects of deflector at the reactor downcomer. In order to investigate the fluid mixing phenomena in the downcomer of an advanced PWR, the flow visualization tests and the salt concentration tests were conducted in a 1/7-scale acrylic transparent model, which was designed and built based on AP 600 reactor geometry. The behaviour of the safety injection flow in downcomer associated with mixing phenomenon can be observed during visualization test, and time-dependent mixing rate between safety injection fluid and existing coolant can be determined with concentration test. Visualization tests were performed by the dye injection method. The results of concentration measurements were compared with the calculation using the REMIX code. During the tests, difference between AP 600 DVI flow and ABB CE System 80+ DVI flow and the effect of the deflector were observed

Performance of Helical Coil Heat Recovery Exchanger using Nanofluid as Coolant

Directory of Open Access Journals (Sweden)

Navid Bozorgan

2015-07-01

Full Text Available Nanofluids are expected to be a promising coolant condidate in chemical processes for heat transfer system size reduction. This paper focuses on reducing the number of turns in a helical coil heat recovery exchanger with a given heat exchange capacity in a biomass heating plant using γ-Al2O3/n-decane nanofluid as coolant. The nanofluid flows through the tubes and the hot n-hexane flows through the shell. The numerical results show that using nanofluid as coolant in a helical coil heat exchanger can reduce the manufacturing cost of the heat exchanger and pumping power by reducing the number of turns of the coil.

Status of zinc injection in PWRs

Energy Technology Data Exchange (ETDEWEB)

Bergmann, C.A. [Westinghouse Electric Co., Pittsburgh, PA (United States)

1995-03-01

Based on laboratory and other studies, it was concluded that zinc addition in a PWR primary coolant should result in reduced Alloy 600 PWSCC and general corrosion rates of the materials of construction. Because of these positive results, a Westinghouse Owner`s Subgroup, EPRI, and Westinghouse provided funds to continue the development and application of zinc in an operating plant. As part of the program, Southern Operating Nuclear Company agreed to operate the Farley 2 plant with zinc addition as a demonstration test of the effectiveness of zinc. Since zinc is incorporated in the corrosion oxide film on the primary system surfaces and Farley 2 is a mature plant, it was estimated that about 10 kgs of zinc would be needed to condition the plant before an equilibrium value in the coolant would be reached. The engineered aspects of a Zinc Addition and Monitoring System (ZAMS) considered such items as the constitutents, location, sizing and water supply of the ZAMS. Baseline data such as the PWSCC history of the Alloy 600 steam generator tubing, fuel oxide thickness, fuel crud deposits, radiation levels, and RCP seal leak-off rates were obtained before zinc addition is initiated. This presentation summarizes some of the work performed under the program, and the status of zinc injection in the Farley 2 plant.

Multirods burst tests under loss-of-coolant conditions

International Nuclear Information System (INIS)

Kawasaki, S.; Uetsuka, H.; Furuta, T.

1983-01-01

In order to know the upper limit of coolant flow area restriction in a fuel assembly under loss-of-coolant accidents in LWRs, burst tests of fuel bundles were performed. Each bundle consisted of 49 rods(7x7 rods), and bursts were conducted in flowing steam. In some cases, 4 rods were replaced by control rods with guide tubes in a bundle. After the burst, the ballooning behavior of each rod and the degree of coolant flow area restriction in the bundle were measured. Ballooning behavior of rods and degree of coolant flow channel restriction in bundles with control rods were not different from those without control rods. The upper limit of coolant flow channel restriction under loss-of-coolant conditions was estimated to be about 80%. (author)

Extended Life Coolant Testing

Science.gov (United States)

2016-06-06

number. PLEASE DO NOT RETURN YOUR FORM TO THE ABOVE ADDRESS. 1. REPORT DATE (DD-MM-YYYY) 06-06-2016 2. REPORT TYPE Interim Report 3. DATES COVERED ... Corrosion Testing of Traditional and Extended Life Coolants 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Hansen, Gregory A. T...providing vehicle specific coolants. Several laboratory corrosion tests were performed according to ASTM D1384 and D2570, but with a 2.5x extended time

Break-up and quench behavior of molten material in coolant

International Nuclear Information System (INIS)

Abe, Y.; Kizu, T.; Arai, T.; Nariai, H.; Chitose, K.; Koyama, K.

2003-01-01

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. The material, being fragmented while solidification and forming debris bed, will be cooled in the coolant. In the experiment, molten material jet is injected into water to experimentally obtain fragments and the visualized information of the fragmentation and boiling phenomena during PAHR in CDA. The distributed particle behavior of the molten material jet is observed with high-speed video camera. The experimental results are compared with the existing theories. Consequently, the marginal wavelength on the surface of a water jet is close to the value estimated based on the Rayleigh-Taylor instability. Moreover, the fragmented droplet diameter obtained from the interaction of molten material and water 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 of the molten particle with some appropriate heat transfer model

SSYST-1. A computer code system to analyse the fuel rod behaviour during a loss of coolant accident

International Nuclear Information System (INIS)

Gulden, W.

1977-08-01

The modules of the SSYST program system allow the detailed analysis of an LWR fuel rod in the course of a postulated loss-of-coolant accident. They provide a tool for considering the interaction between the heat conduction in the fuel rod, heat transfer in the gap, fuel and cladding tube deformation, pressure in the coolant, as well as thermal and fluid dynamics in the cooling channel and for calculating the time and location of ballooning and rod failure, respectively. They can be used both to precalculate the behaviour of fuel rods during LWR accidents and in support of the design of experiments. Depending on the problem to be solved, the individual modules can be easily combined. (orig.) [de

Limits to fuel/coolant mixing

International Nuclear Information System (INIS)

Corradini, M.L.; Moses, G.A.

1985-01-01

The vapor explosion process involves the mixing of fuel with coolant prior to the explosion. A number of analysts have identified limits to the amount of fuel/coolant mixing that could occur within the reactor vessel following a core melt accident. Past models are reviewed and a sim plified approach is suggested to estimate the upper limit on the amount of fuel/coolant mixing pos sible. The approach uses concepts first advanced by Fauske in a different way. The results indicat that water depth is an important parameter as well as the mixing length scale D /SUB mix/ , and for large values of D /SUB mix/ the fuel mass mixed is limited to <7% of the core mass

Enriched boric acid as an optimized neutron absorber in the EPR primary coolant

International Nuclear Information System (INIS)

Cosse, Christelle; Jolivel, Fabienne; Berger, Martial

2012-09-01

reasons, the EBA Boron 10 isotopic atomic abundance target value is 37 at.%. However, EBA conditioning also necessitates the renewal of boric acid supply, operating and management schemes. Supplies must cover the needs for initial conditioning and for later compensation of losses during operation. These losses can be divided into three categories: 1. volumetric losses (such as leaks) 2. decrease of total Boron concentration (dilution for example) 3. Boron neutron core depletion (through Boron 10 consuming reactions). The first two losses are classically compensated with EBA supplied with the required isotopic abundance. Very Enriched boric acid (VEBA) will also be supplied to compensate Boron 10 enrichment losses due to neutron depletion ( 10 B [n, He] 7 Li). According to an optimized boron neutron depletion management scheme, neutron depletion will be directly compensated into boric acid make-up tanks prior to any refuelling shutdown, by injection of reasonable quantities of VEBA (Boron 10 atomic abundance superior to 90 at.%). Finally, EPR operators and chemists will have to cope with these two different sorts of boric acid. Distinctive methods will be employed to avoid any error: use of an optimized and strict policy of different units of Boron used in instrumentation and control systems and in operating procedures; distinct packaging and storage; separate procedures; specific skills for Boron neutron core depletion management. EPR operation has to rely on appropriate parameters and procedures. Total Boron concentration and Boron 10 isotopic abundance monitoring will be periodically achieved by sampling and analysing all borated systems, so called 'boron mapping'. The Boron 10 concentration is continuously monitored in the reactor coolant system with the on-line neutronic boron meter located in the Nuclear Sampling System (NSS): this data is directly used for reactor operation and chemistry control. The use of EBA in the EPR is one of the major progresses for

Development of a reactor-coolant-pump monitoring and diagnostic system. Semi-annual progress report, December 1981-May 1982

International Nuclear Information System (INIS)

Morris, D.J.; Gabler, H.C.

1982-10-01

Reactor coolant (RC) pump seal failures have resulted in excessive leakage of primary coolant into reactor containment buildings. In some cases, high levels of airborne activity and surface contamination following these failures have necessitated extensive cleanup efforts and personnel radiation exposure. Unpredictable pump seal performance has also caused forced outages and frequent maintenance. The quality of operating data has been insufficient to allow proper evaluation of theoretical RC pump seal failure mechanisms. The RC pump monitoring and diagnostic system being developed and installed at Toledo Edison's Davis-Besse Nuclear Power Station will examine the relationship between seal failures and three other variables. This report describes system software and hardware development, testing, and installation work performed during the period of December 1981 through May 1982. Also described herein is a parallel effort being conducted by a B and W/Byron Jackson/Utility group to improve pump seal performance

The PEP injection system

International Nuclear Information System (INIS)

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

1988-01-01

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

Tritium transport modeling at system level for the EUROfusion dual coolant lithium-lead breeding blanket

Science.gov (United States)

Urgorri, F. R.; Moreno, C.; Carella, E.; Rapisarda, D.; Fernández-Berceruelo, I.; Palermo, I.; Ibarra, A.

2017-11-01

The dual coolant lithium lead (DCLL) breeding blanket is one of the four breeder blanket concepts under consideration within the framework of EUROfusion consortium activities. The aim of this work is to develop a model that can dynamically track tritium concentrations and fluxes along each part of the DCLL blanket and the ancillary systems associated to it at any time. Because of tritium nature, the phenomena of diffusion, dissociation, recombination and solubilisation have been modeled in order to describe the interaction between the lead-lithium channels, the structural material, the flow channel inserts and the helium channels that are present in the breeding blanket. Results have been obtained for a pulsed generation scenario for DEMO. The tritium inventory in different parts of the blanket, the permeation rates from the breeder to the secondary coolant and the amount of tritium extracted from the lead-lithium loop have been computed. Results present an oscillating behavior around mean values. The obtained average permeation rate from the liquid metal to the helium is 1.66 mg h-1 while the mean tritium inventory in the whole system is 417 mg. Besides the reference case results, parametric studies of the lead-lithium mass flow rate, the tritium extraction efficiency and the tritium solubility in lead-lithium have been performed showing the reaction of the system to the variation of these parameters.

H- charge exchange injection systems

International Nuclear Information System (INIS)

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

1980-01-01

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

Reactor coolant pump seal leakage monitoring

International Nuclear Information System (INIS)

Stevens, D.M.; Spencer, J.W.; Morris, D.J.; James, W.; Shugars, H.G.

1986-01-01

Problems with reactor coolant pump seals have historically accounted for a large percentage of unscheduled outages. Studies performed for the Electric Power Research Institute (EPRI) have shown that the replacement of coolant pump seals has been one of the leading causes of nuclear plant unavailability over the last ten years. Failures of coolant pump seals can lead to primary coolant leakage rates of 200-500 gallons per minute into the reactor building. Airborne activity and high surface contamination levels following these failures require a major cleanup effort and increases the time and personnel exposure required to refurbish the pump seals. One of the problems in assessing seal integrity is the inability to accurately measure seal leakage. Because seal leakage flow is normally very small, it cannot be sensed directly with normal flow instrumentation, but must be inferred from several other temperature and flow measurements. In operating plants the leakage rate has been quantified with a tipping-bucket gauge, a device which indicates when one quart of water has been accumulated. The tipping-bucket gauge has been used for most rainfall-intensity monitoring. The need for a more accurate and less expensive gauge has been addressed. They have developed a drop-counter precipitation sensor has been developed and optimized. The applicability of the drop-counter device to the problem of measuring seal leakage is being investigated. If a review of system specification and known drop-counter performance indicates that this method is feasible for measuring seal leak rates, a drop-counter gauge will be fabricated and tested in the laboratory. If laboratory tests are successful the gauge will be demonstrated in a pump test loop at Ontario Hydro and evaluated under simulated plant conditions. 3 references, 2 figures

Impedance calculations for power cables to primary coolant pump motors

International Nuclear Information System (INIS)

Hegerhorst, K.B.

1977-01-01

The LOFT primary system motor generator sets are located in Room B-239 and are connected to the primary coolant pumps by means of a power cable. The calculated average impedance of this cable is 0.005323 ohms per unit resistance and 0.006025 ohms per unit reactance based on 369.6 kVA and 480 volts. The report was written to show the development of power cable parameters that are to be used in the SICLOPS (Simulation of LOFT Reactor Coolant Loop Pumping System) digital computer program as written in LTR 1142-16 and also used in the pump coastdowns for the FSAR Analysis

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

Energy Technology Data Exchange (ETDEWEB)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

Reactor coolant pump transportation incident

International Nuclear Information System (INIS)

Noce, D.

1992-01-01

This paper reports on an incident, which occurred on August 27, 1991, in which a Reactor Coolant Pump motor en route from Surry Power Station to Westinghouse repair facilities struck the overpass at the junction of Interstate 64 and Jefferson Avenue in Newport News, Virginia. The transport container that housed the reactor coolant pump motor failed to clear the overpass. The force of the impact dislodged the container and motor from the truck bed, and it landed on the acceleration land and road shoulder. Upon impact, the container broke open and exposed the reactor coolant pump motor. Incidental radioactively contaminated water that remained in the motor coolers drained onto the road, contaminating the aggregate as well as the underlying gravel

Economics of water injected air screw compressor systems

Science.gov (United States)

Venu Madhav, K.; Kovačević, A.

2015-08-01

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

Performance of the ALS injection system

International Nuclear Information System (INIS)

Kim, C.H.

1993-05-01

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

Deposition and incorporation of corrosion product to primary coolant suppressing method

International Nuclear Information System (INIS)

Tsuzuki, Yasuo; Hasegawa, Naoyoshi; Fujioka, Tsunaaki.

1992-01-01

In a PWR type nuclear power plant, the concentration of dissolved nitrogen in primary coolants is increased by controlling the nitrogen partial pressure in a volume controlling tank gas phase portion or addition of water in a primary system water supply tank containing dissolved nitrogen to a primary system. Then ammonium is formed by a reaction with hydrogen dissolved in the primary coolants in the field of radiation rays, to control the concentration of ammonium in the coolants within a range from 0.5 to 3.5 ppm, and operate the power plant. As a result, deposition and incorporation of corrosion products to the structural materials of the primary system equipments during plant operation (pH 6.8 to 8.0) are suppressed. In other words, deposition of particulate corrosion products on the surface of fuel cladding tubes and the inner surface of pipelines in the primary system main equipments is prevented and incorporation of ionic radioactive corrosion products to the oxide membranes on the inner surface of the pipelines of the primary system main equipments is suppressed, to greatly reduce the radiation dose rate of the primary system pipelines. Thus, operator's radiation exposure can be decreased upon shut down of the plant. (N.H.)

Injection system of the minicyclotron accelerator mass spectrometer

International Nuclear Information System (INIS)

Liu Yonghao; Li Deming; Chen Maobai; Lu Xiangshun

1999-01-01

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

Primary coolant circuits in FBR type reactors

International Nuclear Information System (INIS)

Kutani, Masushiro.

1985-01-01

Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

Reliability analysis of the containment spray system of Angra-1 : the injection phase

International Nuclear Information System (INIS)

Gibelli, S.M.O.; Oliveira, L.F.S. de.

1981-12-01

The system studied is projected to perform two basic functions : to reduce the pressure and temperature in the containment after a LOCA (loss of coolant accident), to break the main steam line or the main feed line in the containment after a LOCA (loss of coolant accident), to break the main steam line or the main feed line in the containment and to remove the fission products, mainly the iodine of the containment atmosphere. The spray system was analyzed concerning the probability of non-acomplishment of both functions at the same time; therefore the failure of the components of the chemical aditions subsystem are included in the failure tree shown here. (E.G.) [pt

Operational considerations for the PSB H- Injection System

CERN Document Server

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

2010-01-01

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

Transmutation performance analysis on coolant options in a hybrid reactor system design for high level waste incineration

Energy Technology Data Exchange (ETDEWEB)

Hong, Seong-Hee; Siddique, Muhammad Tariq; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

2015-11-15

Highlights: • Waste transmutation performance was compared and analyzed for seven different coolant options. • Reactions of fission and capture showed big differences depending on coolant options. • Moderation effect significantly affects on energy multiplication, tritium breeding and waste transmutation. • Reduction of radio-toxicities of TRUs showed different trend to coolant choice from performance of waste transmutation. - Abstract: A fusion–fission hybrid reactor (FFHR) is one of the most attractive candidates for high level waste transmutation. The selection of coolant affects the transmutation performance of a FFHR. LiPb coolant, as a conventional coolant for a FFHR, has problems such as reduction in neutron economic and magneto-hydro dynamics (MHD) pressure drop. Therefore, in this work, transmutation performance is evaluated and compared for various coolant options such as LiPb, H{sub 2}O, D{sub 2}O, Na, PbBi, LiF-BeF{sub 2} and NaF-BeF{sub 2} applicable to a hybrid reactor for waste transmutation (Hyb-WT). Design parameters measuring performance of a hybrid reactor were evaluated by MCNPX. They are k{sub eff}, energy multiplication factor, neutron absorption ratio, tritium breeding ratio, waste transmutation ratio, support ratio and radiotoxicity reduction. Compared to LiPb, H{sub 2}O and D{sub 2}O are not suitable for waste transmutation because of neutron moderation effect. Waste transmutation performances with Na and PbBi are similar to each other and not different much from LiPb. Even though molten salt such as LiF-BeF{sub 2} and NaF-BeF{sub 2} is good for avoiding MHD pressure drop problem, waste transmutation performance is dropped compared with LiPb.

Upper internals of PWR with coolant flow separator

International Nuclear Information System (INIS)

Chevereau, G.; Heuze, A.

1989-01-01

The upper internals for a PWR has a collecting volume for the coolant merging from the core and an apparatus for separating the flow of coolant. This apparatus has a guide for the control rods, a lower plate perforated to allow the coolant through from the core, an upper plate also perforated to allow the coolant through to the collecting volume and a peripheral binding ring joining the two plates. Each guide comprises an envelope without holes and joined perceptibly tight to the plates [fr

Interfacing systems LOCAs [Loss of Coolant Accidents] at boiling water reactors

International Nuclear Information System (INIS)

Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

1987-01-01

The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency

Injection system of teh SSC Medium Energy Booster

International Nuclear Information System (INIS)

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

1994-04-01

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

The operating reliability of the reactor coolant pump

International Nuclear Information System (INIS)

Grancy, W.

1996-01-01

There is a strong tendency among operating companies and manufacturers of nuclear power stations to further increase safety and operating availability of the plant and of its components. This applies also and particularly to reactor coolant pumps for the primary circuit of nuclear power stations of the type PWR. For 3 decades, ANDRITZ has developed and built such pumps and has attached great importance to the design of the complete pump rotor and of its essential surrounding elements, such as bearing and shaft seal. Apart from questions connected with design functioning of the pump there is one question of top priority: the operating reliability of the reactor coolant pump. The pump rotor (together with the rotor of the drive motor) is the only component within the primary system that permanently rotates at high speed during operation of the reactor plant. Many questions concerning design and configuration of such components cannot be answered purely theoretically, or they can only be answered partly. Therefore comprehensive development work and testing was necessary to increase the operating reliability of the pump rotor itself and of its surrounding elements. This contribution describes the current status of development and, as a focal point, discusses shaft sealing solutions elaborated so far. In this connection also a sealing system will be presented which aims for the first time at using a two-stage mechanical seal in reactor coolant pumps

Reactor system

International Nuclear Information System (INIS)

Miyano, Hiroshi; Narabayashi, Naoshi.

1990-01-01

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

Numerical analysis of experiments with gas injection into liquid metal coolant

International Nuclear Information System (INIS)

Usov, E V; Lobanov, P D; Pribaturin, N A; Mosunova, N A; Chuhno, V I; Kutlimetov, A E

2016-01-01

Presented paper contains results of a numerical analysis of experiments with gas injection in water and liquid metal which have been performed at the Institute of Thermophysics Russian Academy of Science (IT RAS). Obtained experimental data are very important to predict processes that take place in the BREST-type reactor during the hypothetical accident with damage of the steam generator tubes, and may be used as a benchmark to validate thermo-hydraulic codes. Detailed description of models to simulate transport of gas phase in a vertical liquid column is presented in a current paper. Two-fluid model with closing relation for wall friction and interface friction coefficients was used to simulate processes which take place in a liquid during injection of gaseous phase. It has being shown that proposed models allow obtaining a good agreement between experimental data and calculation results. (paper)

Selection of nuclear reactor coolant materials

International Nuclear Information System (INIS)

Shi Lisheng; Wang Bairong

2012-01-01

Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

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)

IEA-R1 primary and secondary coolant piping systems coupled stress analysis

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A.; Mattar Neto, Miguel

2013-01-01

The aim of this work is to perform the stress analysis of a coupled primary and secondary piping system of the IEA-R1 based on tridimensional model, taking into account the as built conditions. The nuclear research reactor IEA-R1 is a pool type reactor projected by Babcox-Willcox, which is operated by IPEN since 1957. The operation to 5 MW power limit was only possible after the conduction of life management and modernization programs in the last two decades. In these programs the components of the coolant systems, which are responsible for the water circulation into the reactor core to remove the heat generated inside it, were almost totally refurbished. The changes in the primary and secondary systems, mainly the replacement of pump and heat-exchanger, implied in piping layout modifications, and, therefore, the stress condition of the piping systems had to be reanalyzed. In this paper the structural stress assessment of the coupled primary and secondary piping systems is presented and the final results are discussed. (author)

Field experience on Zn injection on PWR plants with a view to dose rate reduction

International Nuclear Information System (INIS)

Roumiguiere, F.

2005-01-01

Operating experience acquired at PWR plants shows that zinc injection in the primary coolant at low concentration (∼5 ppb) is a very effective tool to achieve a reduction of the dose rate build-up. The beneficial effect of zinc consists on improving the protective layer characteristics of the reactor coolant system surfaces, which results in a lower pickup of activated products (Co-60, Co-58), and consequently a reduction of the associated dose rates. Zinc injection was introduced at the Unit B of the Biblis Power Station in September 1996 and at the Obrigheim Nuclear Power Station in February 1998, as a measure for reduction of radiation fields. The effectiveness of the method and its compatibility with the overall plant was examined in a rather comprehensive surveillance program at these plants. The already published data show that zinc injection did not lead to any operating restrictions or other negative effects on plants systems and components. Zinc injection is still being implemented today at these plants. Zinc injection is considered today as a mature technique and is now being successfully applied at a number of PWRs in Germany, Brazil, USA and Japan, with the support of Framatome-ANP. Several PWRs in Europe and Asia are preparing for zinc chemistry in the near future. The method is inexpensive and easy to apply. Its implementation is highly advisable in terms of the cost/benefit criterion following the ALARA principle. This paper gives an overview of the experience gathered with the method. The main subject addressed by the paper is the evolution of dose rates at the primary system and work-related doses since introduction of the method. In German PWRs with Incoloy 800 steam generator tubing material (Ni-content ∼32%), the observed reductions correspond to a decrease in dose rates of around 10 to 15% per year following, as predicted, the half-life time of 60 Co. Overall reductions in high radiation areas are now in the range of 50% after 5 years of

Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

Energy Technology Data Exchange (ETDEWEB)

Elina Syrjaelahti; Anitta Haemaelaeinen [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)

2005-07-01

Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

International Nuclear Information System (INIS)

Elina Syrjaelahti; Anitta Haemaelaeinen

2005-01-01

Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

Primary Coolant pH Control for Soluble Boron-Free PWRs

International Nuclear Information System (INIS)

Cheon, Yang Ho; Lee, Nam Yeong; Park, Byeong Ho; Park, Seong Chan; Kim, Eun Kee

2015-01-01

These should be considered when evaluating and designing the operating pH program for nuclear power plants. This paper discusses the advanced water chemistry strategies to keep pace with the recent global trends related to pH control in the primary water system for soluble boron pressurized water reactor (PWR) plants. Finally, the objective of this work is to study primary coolant pH control for soluble boron-free PWR plants. This paper reviewed the advanced water chemistry strategies to keep pace with the recent global trends related to pH control in the primary water chemistry system for soluble boron PWR plants. The new chemistry trend for the primary coolant is towards adaption of the constant and elevated chemistry. Finally, this work studied primary coolant pH control for soluble boron-free PWR plants. The ammonia-based water chemistry related to pH control for boron-free PWR plants was discussed. The ammonia-based water chemistry is not recommended to avoid fluctuation of the pH value by ammonia radiolysis and to reduce C-14 production in reactor coolant from reaction with dissolved nitrogen. Also, the potassium-based water chemistry related to pH control for boron-free PWR plants was discussed. KOH has a potential as an alternative pH control agent for soluble boron-free PWR plants. The potassium-based water chemistry related to pH control is recommended for boron-free operation as follows

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.)

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.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

EDF PWRs primary coolant purification strategies

International Nuclear Information System (INIS)

Gressier, Frederic; Mascarenhas, Darren; Taunier, Stephane; Le-Calvar, Marc; Bretelle, Jean-Luc; Ranchoux, Gilles

2012-09-01

In order to achieve a good physico-chemical quality of the primary coolant fluid, the primary water is continuously treated by the Chemical and Volume Control System (CVCS). This system is composed of a treatment chain containing filters and ion-exchange resins. In the EDF design, an upstream filter is placed before the resin so as to prevent it from being saturated with insoluble particles. Then, the fluid passes through several resin beds (up to 3 depending on the configuration) and again through a downstream filter that prevents resin fines dissemination into the reactor coolant. Much work has been conducted in the last 5 years on the homogenisation of products and usage on French EDF NPP primary coolant treatment, while taking into account the compromise between source term reduction, liquid and solid waste, and buying and disposal costs. Two national markets have been created, and two operational documents for chemists on site have been published: a filtration guideline and an ion-exchange resin guideline. Both documents give general information about the products used, how are they characterized and selected for national market (technical requirements, standards and tests), how they should be used and what are the change-out criteria. They are also periodically updated based on feedback from sites. The positive impact on resin and filter lifetime (extension of some, limitation of others), homogenisation of products and usage will be presented. Moreover, EDF is constantly in the process of improving the current purification methods, as well as researching the use of existing and novel technologies. In this field, recent experiments on short loading of resin during reactor shutdown has been tested on site with success. In addition, work is done on silica free filters, filter consumption and filter chemical release. An overview of these optimization methods will be given. (authors)

The measurement of moisture in nuclear coolant gases. Experience and new developments within the CEGB

International Nuclear Information System (INIS)

Hiorns, D.S.; Stallard, M.D.

1982-06-01

Humidity measurements on nuclear reactor coolant gases are required for many reasons, for instance to detect and locate water leakages from the boilers and to monitor the chemical composition of the coolant for normal operation and during commissioning and maintenance, and the assessment of the performance of driers. For leak detection sensitivity and rapid response are important whereas for routine monitoring drift may be a prime criterion. Some coolants include methane for chemical control that decomposes to water so the levels of moisture that normally exist in the coolant vary with reactor types. No one measurement technique or hygrometer system currently available can be expected to meet all these requirements and in the CEGB it has proved necessary to concentrate on finding a solution to each separately. This paper reviews CEGB experience with numerous hygrometer systems that have been used on both Magnox and AGR coolant gases and highlights some of the problems that have arisen. New development work in the important and demanding area of boiler leak detection and location is also included where a number of new approaches are being investigated, including:- (i) An improved 'first-up' system based on coulometric moisture analysers. (ii) A differential system based on the piezo-electric hygrometer. (iii) A microprocessor controlled system using capacitance probes. Whilst work on the development of new instruments and the design of approved systems is proceeding at different locations within the CEGB it is being reported, together with the latest operational experiences, through the CEGB Chemical Measuring Instruments Advisory Group, on whose behalf this paper is presented. (author)

Behaviour analysis of the fuel injected in the intake manifold of port-injected spark ignition engines: modeling and experimental validation; Analyse du comportement du carburant injecte dans les conduits d`admission des moteurs a allumage commande a injection multipoint: modelisation et validation experimentale

Energy Technology Data Exchange (ETDEWEB)

Sches, C

1999-01-27

In order to limit pollutant emissions resulting from transient engine operation, the mastering of mixture formation is essential. In this context, an interactive work was undertaken between a modeling job and an experimental study, to get better understanding of the mechanisms of fuel dynamic behavior in the intake manifold of port-injected spark-ignition engines. The experimental study, elaborated thanks to experimental designs, showed out two essential factors: injection timing and coolant liquid temperature, which act on the fuel dynamic behavior through a second order filter. Then, a phenomenological modeling was established and validated, to analyze the various phenomena influencing mixture formation and to calculate the air/fuel ratio evolutions during transient operation. This program uses the results of a 3D model describing the fuel spray transportation, evaporation and impact on the port walls. The calculation does not need any boundary conditions and the running times are vary satisfactory. We showed that a correct description of the liquid fuel film was necessary to get good prediction of the mixture fuel/air ratio. The spray modeling, which is necessary, can however be kept simple. Future work may develop either in the engine control filed (injection strategies development, optimization of the injection system configuration, ...), or in the theoretical field (better modeling of fuel film displacement or of secondary atomization of the fuel on the intake valve). (author) 79 refs.

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)

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

International Nuclear Information System (INIS)

Boudy, Frederic; Seers, Patrice

2009-01-01

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

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

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

Optimum design of the injection duct system of a stenter machine

Energy Technology Data Exchange (ETDEWEB)

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

2017-05-15

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

The 1994 loss of coolant incident at Pickering NGS

Energy Technology Data Exchange (ETDEWEB)

Charlebois, P R; Clarke, T R; Goodman, R M; McEwan, W F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station; Cuttler, J M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

Fracture of the rubber diaphragm in a liquid relief valve initiated events leading to a loss of coolant in Unit 2, on December 10. The valve failed open, filling the bleed condenser. The reactor shut itself down. When pressure recovered, two spring-loaded safety relief valves opened and one of them chattered. The shock and pulsations cracked the inlet pipe to the chattering valve, and the subsequent loss of coolant triggered the emergency core cooling system. The incident was terminated by operator action. No abnormal radioactivity was released. The four reactor units of Pickering A remained shut down until the corrective actions were completed in April/May 1995. (author). 4 figs.

Loss-of-Coolant and Loss-of-Flow Accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-01-01

This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

Loss-of-coolant and loss-of-flow accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1994-07-01

This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

International Nuclear Information System (INIS)

Chen, N.C.J.; Yoder, G.L.; Wendel, M.W.

1991-01-01

Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs

Reactor Primary Coolant System Pipe Rupture Study. Progress report No. 32, July--December 1974

International Nuclear Information System (INIS)

1975-03-01

The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase I), analytical and experimental efforts (Phase II) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue studies focused on Elastic/Plastic ASME Code Design Rules, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, and (c) studies directed at quantifying weld sensitization in T-304 stainless steel. (auth)

Decontamination of main coolant pumps

International Nuclear Information System (INIS)

Roofthooft, R.

1988-01-01

Last year a number of main coolant pumps in Belgian nuclear power plants were decontaminated. A new method has been developed to reduce the time taken for decontamination and the volume of waste to be treated. The method comprises two phases: Oxidation with permanganate in nitric acid and dissolution in oxalic acid. The decontamination of main coolant pumps can now be achieved in less than one day. The decontamination factors attained range between 15 and 150. (orig.) [de

Effects of Coolant Temperature Changes on Reactivity for Various Coolants in a Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

International Nuclear Information System (INIS)

Casino, William A. Jr.

2006-01-01

The purpose of this study is to perform an investigation into the relative merit of various salts and salt compounds being considered for use as coolants in the liquid salt cooled very high temperature reactor platform (LS-VHTR). Most of the non-nuclear properties necessary to evaluate these salts are known, but the neutronic characteristics important to reactor core design are still in need of a more extensive examination. This report provides a two-fold approach to further this investigation. First, a list of qualifying salts is assembled based upon acceptable non-nuclear properties. Second, the effect on system reactivity for a secondary system transient or an off-normal or accident condition is examined for each of these salt choices. The specific incident to be investigated is an increase in primary coolant temperature beyond normal operating parameters. In order to perform the relative merit comparison of each candidate salt, the System Temperature Coefficient of Reactivity is calculated for each candidate salt at various state points throughout the core burn history. (author)

Mathematical Model-Based Temperature Preparation of Liquid-Propellant Components Cooled by Liquid Nitrogen in the Heat Exchanger with a Coolant

Directory of Open Access Journals (Sweden)

S. K. Pavlov

2014-01-01

Full Text Available Before fuelling the tanks of missiles, boosters, and spacecraft with liquid-propellant components (LPC their temperature preparation is needed. The missile-system ground equipment performs this operation during prelaunch processing of space-purpose missiles (SPM. Usually, the fuel cooling is necessary to increase its density and provide heat compensation during prelaunch operation of SPM. The fuel temperature control systems (FTCS using different principles of operation and types of coolants are applied for fuel cooling.To determine parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is cooled by liquid nitrogen upon contact heat exchange in the coolant reservoir, a mathematical model of this process and a design technique are necessary. Both allow us to determine design parameters of the cooling system and the required liquid nitrogen reserve to cool LPC to the appropriate temperature.The article presents an overview of foreign and domestic publications on cooling processes research and implementation using cryogenic products such as liquid nitrogen. The article draws a conclusion that it is necessary to determine the parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is liquid nitrogen-cooled upon contact heat exchange in the coolant reservoir allowing to define rational propellant cooling conditions to the specified temperature.The mathematical model describes the set task on the assumption that a heat exchange between the LPC and the coolant in the heat exchanger and with the environment through the walls of tanks and pipelines of circulation loops is quasi-stationary.The obtained curves allow us to calculate temperature changes of LPC and coolant, cooling time and liquid nitrogen consumption, depending on the process parameters such as a flow rate of liquid nitrogen, initial coolant temperature, pump characteristics, thermal

Passive safety injection experiments and analyses (PAHKO)

International Nuclear Information System (INIS)

Tuunanen, J.

1998-01-01

PAHKO project involved experiments on the PACTEL facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines (Pressure Balancing Line, PBL, and Injection Line, IL). The examined PSIS worked efficiently in SBLOCAs although the flow through the PSIS stopped temporarily if the break was very small and the hot water filled the CMT. The experiments demonstrated the importance of the flow distributor in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable to simulate the overall behaviour of the transients. The detailed analyses of the results showed some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of the PSIS phenomena. (orig.)

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.)

Liquid metal coolant disposal from UKAEA reactors at Dounreay

International Nuclear Information System (INIS)

Adam, E.R.

1997-01-01

As part of the United Kingdom's Fast Reactor Development programme two reactors were built and operated at Dounreay in the North of Scotland. DFR (Dounreay Fast Reactor) was operated from 1959-1977 and PFR (Prototype Fast Reactor) was operated from 1974-1994. Both reactors are currently undergoing Stage 1 Decommissioning and are installing plant to dispose of the bulk coolant (DFR ∼ 60 tonne; PFR ∼ 1500 tonne). The coolant (NaK) remaining at DFR is mainly in the primary circuit which contains in excess of 500 TBq of Cs137. Disposal of 40 tonnes of secondary coolant has already been carried out. The paper will describe the processes used to dispose of this secondary circuit coolant and how it is intended the remaining primary circuit coolant will be handled. The programme to process the primary coolant will also be described which involves the conversion of the liquid metal to caustic and its decontamination. No PFR coolant Na has been disposed off to date. The paper will describe the current decommissioning programme activities relating to liquid metal disposal and treatment describing the materials to be disposed of and the issue of decontamination of the effluents. (author)

Research on physical and chemical parameters of coolant in Light-Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

2015-07-01

The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

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

Condition monitoring of main coolant pumps, Dhruva

International Nuclear Information System (INIS)

Prasad, V.; Satheesh, C.; Acharya, V.N.; Tikku, A.C.; Mishra, S.K.

2002-01-01

Full text: Dhruva is a 100 MW research reactor with natural uranium fuel, heavy water as moderator and primary coolant. Three Centrifugal pumps circulate the primary coolant across the core and the heat exchangers. Each pump is coupled to a flywheel (FW) assembly in order to meet operational safety requirements. All the 3 main coolant pump (MCP) sets are required to operate during operation of the reactor. The pump-sets are in operation since the year 1984 and have logged more than 1,00,000 hrs. Frequent breakdowns of its FW bearings were experienced during initial years of operation. Condition monitoring of these pumps, largely on vibration based parameters, was initiated on regular basis. Break-downs of main coolant pumps reduced considerably due to the fair accurate predictions of incipient break-downs and timely maintenance efforts. An effort is made in this paper to share the experience

Environmental Assisted Fatigue Evaluation of Direct Vessel Injection Piping Considering Thermal Stratification

International Nuclear Information System (INIS)

Kim, Taesoon; Lee, Dohwan

2016-01-01

As the environmentally assisted fatigue (EAF) due to the primary water conditions is to be a critical issue, the fatigue evaluation for the components and pipes exposed to light water reactor coolant conditions has become increasingly important. Therefore, many studies to evaluate the fatigue life of the components and pipes in LWR coolant environments on fatigue life of materials have been conducted. Among many components and pipes of nuclear power plants, the direct vessel injection piping is known to one of the most vulnerable pipe systems because of thermal stratification occurred in that systems. Thermal stratification occurs because the density of water changes significantly with temperature. In this study, fatigue analysis for DVI piping using finite element analysis has been conducted and those results showed that the results met design conditions related with the environmental fatigue evaluation of safety class 1 pipes in nuclear power plants. Structural and fatigue integrity for the DVI piping system that thermal stratification occurred during the plant operation has conducted. First of all, thermal distribution of the piping system is calculated by computational fluid dynamic analysis to analyze the structural integrity of that piping system. And the fatigue life evaluation considering environmental effects was carried out. Our results showed that the DVI piping system had enough structural integrity and fatigue life during the design lifetime of 60 years

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

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

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.

Distribution and behavior of tritium in the Coolant-Salt Technology Facility

International Nuclear Information System (INIS)

Mays, G.T.; Smith, A.N.; Engel, J.R.

1977-04-01

A 1000-MW(e) Molten-Salt Breeder Reactor (MSBR) is expected to produce 2420 Ci/day of tritium. As much as 60 percent of the tritium produced may be transported to the reactor steam system (assuming no retention by the secondary coolant salt), where it would be released to the environment. Such a release rate would be unacceptable. Experiments were conducted in an engineering-scale facility--the Coolant-Salt Technology Facility (CSTF)--to examine the potential of sodium fluoroborate, the proposed coolant salt for an MSBR, for sequestering tritium. The salt was believed to contain chemical species capable of trapping tritium. A series of 5 experiments--3 transient and 2 steady-state experiments--was conducted from July of 1975 through June of 1976 where tritium was added to the CSTF. The CSTF circulated sodium fluoroborate at temperatures and pressures typical of MSBR operating conditions. Results from the experiments indicated that over 90 percent of tritium added at steady-state conditions was trapped by sodium fluoroborate and appeared in the off-gas system in a chemically combined (water-soluble) form and that a total of approximately 98 percent of the tritium added at steady-state conditions was removed through the off-gas system overall

Water electrolysis system

International Nuclear Information System (INIS)

Mizoguchi, Tadao; Ikehara, Masahisa; Kataoka, Noboru; Ueno, Syuichi; Ishikawa, Nobuhide.

1996-01-01

Nissho Iwai Co. and Ebara Co. received an order for hydrogen and oxygen generating system (water electrolysis system) to be installed at Tokai-2 power station of The Japan Atomic Power Company, following the previous order at Tsuruga-1 where the gas injection from FY1996 is planned. Hydrogen gas generated by the system will be injected to coolant of boiling water reactors to improve corrosive environment. The system is being offered by a tripartite party, Nissho Iwai, Ebara, and Norsk Hydro Electrolysers of Norway (NHEL). NHEL provides a electrolyser unit, as a core of the system. Ebara provides procurement, installation, and inspection as well as total engineering work, under the basic design by NHEL which has over 60 years-experience in this field. (author)

A new dual injection system for AMS facility

International Nuclear Information System (INIS)

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

2007-01-01

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

Dimethyl Ether in Diesel Fuel Injection Systems

DEFF Research Database (Denmark)

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

1998-01-01

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

Los Alamos Proton Storage Ring (PSR) injection deflector system

International Nuclear Information System (INIS)

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

1983-01-01

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

Transient Temperature Distribution in a Reactor Core with Cylindrical Fuel Rods and Compressible Coolant

Energy Technology Data Exchange (ETDEWEB)

Vollmer, H

1968-04-15

Applying linearization and Laplace transformation the transient temperature distribution and weighted temperatures in fuel, canning and coolant are calculated analytically in two-dimensional cylindrical geometry for constant material properties in fuel and canning. The model to be presented includes previous models as special cases and has the following novel features: compressibility of the coolant is accounted for. The material properties of the coolant are variable. All quantities determining the temperature field are taken into account. It is shown that the solution for fuel and canning temperature may be given by the aid of 4 basic transfer functions depending on only two variables. These functions are calculated for all relevant rod geometries and material constants. The integrals involved in transfer functions determining coolant temperatures are solved for the most part generally by application of coordinate and Laplace transformation. The model was originally developed for use in steam cooled fast reactor analysis where the coolant temperature rise and compressibility are considerable. It may be applied to other fast or thermal systems after suitable simplifications.

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)

Analysis of small break loss of coolant accident for Chinese CPR1000

Energy Technology Data Exchange (ETDEWEB)

Kim, Ju Youl [FNC Technology Co., Yongin (Korea, Republic of); Cilier, Anthonie [North-West University, Mahikeng (South Africa); Poc, Li-chi Cliff [Micro-Simulation Technology, Montville (United States)

2016-05-15

This research analyses the small break loss of coolant accident (LOCA) on a Chinese CPR1000 type reactor. LOCA accident is used as benchmark for the PCTRAN/CPR1000 code by comparing the effects and results to the Manshaan FSAR accident analysis. LOCA is a design basis accident in which a guillotine break is postulated to occur in one of the cold legs of a pressurized water reactor (PWR). Consequently, the primary system pressure would drop and almost all the reactor coolant would be discharged into the reactor containment. The drop in pressure would activate the reactor protection system and the reactor would trip. The simulation of a 3-inch small break loss of coolant accident using the PCTRAN/CPR1000 has revealed this code's effectiveness as well as weaknesses in specific simulation applications. The code has the ability to run at 16 times real time and produce very accurate results. The results are consistently producing the same trends as licensed codes used in Safety Assessment Reports. It is however able to produce these results in a fraction of the time and also provides a whole plant simulation coupling the various thermal, hydraulic, chemical and neutronic systems together with a plant specific control system.

PLT neutral beam injection systems

International Nuclear Information System (INIS)

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

1979-01-01

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

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%

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

RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

International Nuclear Information System (INIS)

Kawai, Hiroshi

2002-01-01

In the Chemical Volume Control System (CVCS) reactor primary coolant leakage incident, which occurred in Tsuruga-2 (4-loop PWR, 3,423 MWt, 1,160 MWe) on July 12, 1999, it took about 14 hours before the leakage isolation. The delayed leakage isolation and a large amount of leakage have become a social concern. Effective procedure modification was studied. Three betterments were proposed based on a qualitative analysis to reduce the pressure and temperature of the primary loop as fast as possible by the current plant facilities while maintaining enough subcooling of the primary loop. I analyzed the incident with RETRAN code in order to quantitatively evaluate the leakage reduction when these betterments are adopted. This paper is very new because it created a typical analysis method for PWR plant behavior during plant shutdown procedure which conventional RETRAN transient analyses rarely dealt with. Also the event time is very long. To carry out this analysis successfully, I devised new models such as an Residual Heat Removal System (RHR) model etc. and simplified parts of the conventional model. Based on the analysis results, I confirmed that leakage can be reduced by about 30% by adopting these betterments. Then the Japan Atomic Power Company (JAPC) modified the operational procedure for reactor primary coolant leakage events adopting these betterments. (author)

Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5; Simulacion de un escenario de perdida de refrigerante grande (LBLOCA), sin actuacion de los sistemas de inyeccion de emergencia (ECCS) para un reactor BWR-5

Energy Technology Data Exchange (ETDEWEB)

Cardenas V, J.; Mugica R, C. A.; Lopez M, R., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

2015-09-15

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{sup 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)

Design of a molten heavy-metal coolant and target for fast-thermal accelerator driven sub-critical system (ADS)

International Nuclear Information System (INIS)

Satyamurthy, P.; Degwekar, S.B.; Nema, P.K.

2001-01-01

Accelerator Driven sub-critical Systems (ADS) have evoked considerable interest in recent years. The Energy Amplifier concept developed by C. Rubbia and others at CERN incorporates a buoyancy driven, lead-coolant primary system for extracting the heat generated in the fast reactor as well as that in neutron spallation target. In earlier publications, our BARC group has proposed a one-way coupled booster reactor system which could be operated at proton beam currents as low as 1-2 mA for a power output of 750 MW th . Here, the basic idea is to have a fast booster reactor zone of low power (- 100 MW th ) which is separated by a large gap from the main thermal reactor zone. In this arrangement, the spallation neutron source feeds neutrons to the fast reactor zone where neutrons are further multiplied. Further in this system, the neutrons from the booster region enter the main reactor but very few neutrons from main reactor return to booster, thus ensuring one-way coupling. In earlier work, several possible configurations of the booster and thermal regions were presented. In the present work, we describe an engineering design particularly with respect to thermal hydraulics of lead/lead-bismuth eutectic coolant also acting as spallation neutron source. This hybrid ADS reactor consists of fast and thermal reactor zones producing about 100 MW th and 650 MW th respectively. The scheme of the system is shown. The fast core consists of 48 hexagonal fuel bundles each containing 169 fuel pins of 8.2 mm diameter arranged in 11.4 mm triangular array pitch. The average thermal power per fuel pin is about 13.46 kw. However, due to neutron flux peaking effect, the maximum fuel pin power can be up to 2.5 times this average power. The thermal reactor consists of heavy water as moderator and coolant similar to a typical CANDU type Indian PHWR except for fuel composition. Though the gap between fast and thermal zones essentially provides one way coupling of neutron flux, a thermal

New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

Science.gov (United States)

Rezania, A.; Rosendahl, L. A.

2012-06-01

The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The three-dimensional governing equations for the fluid flow and the heat transfer are solved using the finite-volume method for a wide range of pressure drop laminar flows along the heat sink. The temperature and the mass flow rate distribution in the heat sink are discussed. The results, which are in good agreement with previous computational studies, show that using suggested heat sink configurations reduces the coolant pumping power in the system.

Zero waste machine coolant management strategy at Los Alamos National Laboratory

International Nuclear Information System (INIS)

Carlson, B.; Algarra, F.; Wilburn, D.

1998-01-01

Machine coolants are used in machining equipment including lathes, grinders, saws and drills. The purpose of coolants is to wash away machinery debris in the form of metal fines, lubricate, and disperse heat between the part and the machine tool. An effective coolant prolongs tool life and protects against part rejection, commonly due to scoring or scorching. Traditionally, coolants have a very short effective life in the machine, often times being disposed of as frequently as once per week. The cause of coolant degradation is primarily due to the effects of bacteria, which thrive in the organic rich coolant environment. Bacteria in this environment reproduce at a logarithmic rate, destroying the coolant desirable aspects and causing potential worker health risks associated with the use of biocides to control the bacteria. The strategy described in this paper has effectively controlled bacterial activity without the use of biocides, avoided disposal of a hazardous waste, and has extended coolant life indefinitely. The Machine Coolant Management Strategy employed a combination of filtration, heavy lubricating oil removal, and aeration, which maintained the coolant peak performance without the use of biocides. In FY96, the Laboratory generated and disposed of 19,880 kg of coolants from 9 separate sites at a cost of $145K. The single largest generator was the main machine shop producing an average 14,000 kg annually. However, in FY97, the waste generation for the main machine shop dropped to 4,000 kg after the implementation of the zero waste strategy. It is expected that this value will be further reduced in FY98

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

Science.gov (United States)

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

2010-01-01

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

Exhaust temperature analysis of four stroke diesel engine by using MWCNT/Water nanofluids as coolant

Science.gov (United States)

Muruganandam, M.; Mukesh Kumar, P. C.

2017-10-01

There has been a continuous improvement in designing of cooling system and in quality of internal combustion engine coolants. The liquid engine coolant used in early days faced many difficulties such as low boiling, freezing points and inherently poor thermal conductivity. Moreover, the conventional coolants have reached their limitations of heat dissipating capacity. New heat transfer fluids have been developed and named as nanofluids to try to replace traditional coolants. Moreover, many works are going on the application of nanofluids to avail the benefits of them. In this experimental investigation, 0.1, 0.3 and 0.5% volume concentrations of multi walled carbon nanotube (MWCNT)/water nanofluids have been prepared by two step method with surfactant and is used as a coolant in four stroke single cylinder diesel engine to assess the exhaust temperature of the engine. The nanofluid prepared is characterized with scanning electron microscope (SEM) to confirm uniform dispersion and stability of nanotube with zeta potential analyzer. Experimental tests are performed by various mass flow rate such as 270 300 330 LPH (litre per hour) of coolant nanofluids and by changing the load in the range of 0 to 2000 W and by keeping the engine speed constant. It is found that the exhaust temperature decreases by 10-20% when compared to water as coolant at the same condition.

Piezoelectric Injection Systems

Science.gov (United States)

Mock, R.; Lubitz, K.

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

Fuel-coolant interactions: preliminary experiments on the effect of gases dissolved in the 'coolant'

International Nuclear Information System (INIS)

Asher, R.C.; Davies, D.; Jones, P.G.

1976-12-01

A simple apparatus has been used to study fuel-coolant interactions under reasonably well controlled conditions. Preliminary experiments have used water as the 'coolant' and molten tin at 800 0 C as the 'fuel' and have investigated how the violence of the interaction is affected by dissolving gases (oxygen, nitrogen, carbon dioxide and nitrous oxide) in the water. It was found that saturating the water with carbon dioxide or nitrous oxide completely suppresses the violent interaction. Experiments in which the concentrations of these gases were varied showed that a certain critical concentration was needed; below this concentration the dissolved gas has no significant effect but above it the suppression is

Nuclear reactor of pressurized liquid coolant type

International Nuclear Information System (INIS)

Costes, D.

1976-01-01

The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

Use of a multi-attribute utility theory for evaluating the best coolant material in transmutation reactors

International Nuclear Information System (INIS)

Yu, Dong Han; Han, Suk Joong; Kim, Do Hyung; Park, Won Suk

1998-12-01

In order to develop and design a good transmutation system, it is necessary first to select the best available coolant material for a reactor coolant system. Choosing the best coolant material may not be easy since there are several criteria associated with thermal performance, safety problem, cost problem, neutronic aspects. etc. The best option should be chosen based on the maximization of our needs in this situation. It is a challenging task. Decision theory can be employed to solve this type of problem. This report presents the feasibility study for evaluating the best coolant material in transmutation reactors based on the multi=attribute utility theory. The main problem presented here is how to logically evaluate candidate coolant materials under multiple criteria such as thermal performance, safety problem, cost problem, cost problem, neutronic aspects, etc. Since the current problem involves multiple criteria or attributes, first of all, the multi-attribute utility theory (MAUT) such as SMART and AHP has been extensively reviewed. Then, many candidate coolant material for transmutation reactors have been identified. The next step is to construct a value tree that express to reflect the relative importance of the attributes for overall evaluation. Finally, given these assignments, the final goal were obtained by manipulating these ranks through the value tree. The proposed approach is intended to help people be rational and logical in making decisions such complex tasks. (author). 8 refs., 7 tabs., 22 figs

Steam as turbine blade coolant: Experimental data generation

Energy Technology Data Exchange (ETDEWEB)

Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

1995-10-01

Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

Assessment of the integration of a He-cooled divertor system in the power conversion system for the dual-coolant blanket concept (TW2-TRP-PPCS12D8)

International Nuclear Information System (INIS)

Norajitra, P.; Kruessmann, R.; Malang, S.; Reimann, G.

2002-12-01

Application of a helium-cooled divertor together with the dual-coolant blanket concept is considered favourable for achieving a high thermal efficiency of the power plant due to its relatively high coolant outlet temperature. A new FZK He-cooled modular divertor concept with integrated pin arrays (HEMP) is introduced. Its main features and function are described in detail. The result of the thermalhydraulic analysis shows that the HEMP divertor concept has the potential of resisting, a heat flow density of at least 10-15 MW/m 2 at a reachable heat transfer coefficient of approx. 60 kW/m 2 K and a reasonable pumping power. Integration of this divertor concept into the power conversion system using a closed Brayton gas turbine system with three-stage compression leads to a net efficiency of the blanket/divertor cycle of about 43%. (orig.)

Methodologies and technologies for life assessment and management of coolant channels of Indian pressurised heavy water reactors

International Nuclear Information System (INIS)

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

2002-01-01

coolant channels and various research publications in international journals formed the bases for evolution of the above safety methodologies. Today, the analytical models together with the safety evaluation methodologies have become an important tool for assessing fitness for service of individual pressure tube in different reactors. Apart from developing the above analytical tools and methodologies for safety evaluation, several systems and tools have been designed and developed to cater for the activities like remote inspection, online integrity monitoring and life extension etc., of our life management programme of coolant channel. Amongst the systems developed for remote inspection of the coolant channel such as Dry Visual Inspection System (DRYVIS), BARC Inspection System(BARCIS) and Hydraulic Remote Inside Diameter Measurement System (HYRIM) etc., BARCIS is more versatile in terms of its capability to carry out different type of measurements. Inspection of the coolant channel by BARCIS has been incorporated as an important activity in our life management programme of the coolant channel. A non destructive diagnostic technique based on the principle of vibration measurement has been developed to identify the contact between the pressure tube and calandria tube. This technique called 'Non Intrusive Vibration Diagnostic Technique (NIVDT)' is being used as a tool to screen the vulnerable pressure tubes so as reduce the reduce the inspection load. Tools and techniques developed for online integrity monitoring includes Sliver Sample Scraping Tool (SSST). At present, scraping operation can be carried out in both the dry and the wet channel of operating reactor. The development of SSST is intended for taking nondestructively sliver samples from the operating coolant channel. These sliver samples are then analysed to get hydrogen content in them. Measured hydrogen content helps not only in predicting the operating life of a coolant channel but also in improving the

Dual coolant blanket concept

International Nuclear Information System (INIS)

Malang, S.; Schleisiek, K.

1994-11-01

A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

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

Theoretical studying the stability of steady-state regime of a channel with a coolant condensation

International Nuclear Information System (INIS)

Savikhin, O.G.

1987-01-01

Based on the boiling channel stability theory, the channel steady-state stability with the coolant condensation is studied. Condensable coolants are used in the NPP steam-separator superheaters as well as in cryogenic technique. Under certain conditions the coolant flow rate and temperature fluctuations may be excited in the parallel channel system with coolant condensation, which produce a sufficient effect on the heat exchange equipment operation reliability. To describe unsteady processes of heat and mass transfer in the channel, a homogeneous two-phase flow one dimensional model is used. The results obtained allow one to make a conclusion concerning the effect of some parameters on condensing channel steady-state regime stability: reduction of inlet and outlet unheated communication length, pressure drop increase at the outlet plate and its reduction at the inlet one lead to the increase of stability margin

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

DIAGNOSTICS OF GASOLINE FUEL SYSTEMS WITH DIRECT INJECTION

Directory of Open Access Journals (Sweden)

M. Bulgakov

2017-11-01

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

Simulation of steam explosion in stratified melt-coolant configuration

International Nuclear Information System (INIS)

Leskovar, Matjaž; Centrih, Vasilij; Uršič, Mitja

2016-01-01

Highlights: • Strong steam explosions may develop spontaneously in stratified configurations. • Considerable melt-coolant premixed layer formed in subcooled water with hot melts. • Analysis with MC3D code provided insight into stratified steam explosion phenomenon. • Up to 25% of poured melt was mixed with water and available for steam explosion. • Better instrumented experiments needed to determine dominant mixing process. - Abstract: A steam explosion is an energetic fuel coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations where sufficiently deep coolant pool conditions provide complete jet breakup and efficient premixture formation. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. Based on the hypothesis that there are no interfacial instabilities in a stratified configuration it was assumed that the amount of melt in the premixture is insufficient to produce strong explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts and subcooled water conditions a considerable melt-coolant premixed layer is formed. In the article, the performed study of steam explosions in a stratified melt-coolant configuration in PULiMS like conditions is presented. The goal of this analytical work is to supplement the experimental activities within the PULiMS research program by addressing the key questions, especially regarding the explosivity of the formed premixed layer and the mechanisms responsible for the melt-water mixing. To

Thermal Behavior of the Coolant in the Emergency Cooldown Tank for an Integral Reactor

Energy Technology Data Exchange (ETDEWEB)

Moon, Joo Hyung; Kim, Seok; Kim, Woo Shik; Jung, Seo Yoon; Kim, Young In [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The Residual Heat Removal System (PRHRS) is one of the passive safety systems which should be activated after an accident to remove the residual heat from the core and the sensible heat of the reactor coolant system (RCS) through the steam generators until the safe shutdown conditions are reached. In the previous study presented at the last KNS Autumn Meeting, transient behavior of the RCS temperature and the cooling performance of the PRHRS were investigated numerically by using newly developed in-house code based on MATLAB software. By using the program, the steady-state and transient (quasi-steady state) characteristics during the operation of the PRHRS had been reported. In this program, the temperature of the coolant in the Emergency Cooldown Tank (ECT) was assumed to be constant at saturated state and pool boiling heat transfer mechanism was applied through the entire time domain. The coolant of the ECT reached at a saturated state in early time. It was revealed that the assumption made in the previous study was reasonable.

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

The kicker magnet system for TRISTAN Accumulation Ring injection

International Nuclear Information System (INIS)

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

1994-12-01

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

Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

International Nuclear Information System (INIS)

Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

2000-01-01

Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

Compact toroid injection system for JFT-2M

Energy Technology Data Exchange (ETDEWEB)

Fukumoto, N. [University of Hyogo, 2167 Shosha, Himeji, Hyogo 671-2280 (Japan)]. E-mail: fukumotn@eng.u-hyogo.ac.jp; Ogawa, H. [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Nagata, M. [University of Hyogo, 2167 Shosha, Himeji, Hyogo 671-2280 (Japan); Uyama, T. [University of Hyogo, 2167 Shosha, Himeji, Hyogo 671-2280 (Japan); Shibata, T. [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Kashiwa, Y. [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Suzuki, S. [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Kusama, Y. [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2006-11-15

The compact toroid (CT) injection system for JFT-2M is composed of a CT injector, a gas delivery and vacuum system, a power supply system, and a diagnostics system. In particular, the power supply system delivers high performance for CT formation and acceleration. The CT formation capacitor bank unit achieved a formation current of 350 kA with a rise time less than 10 {mu}s. Although the CT acceleration bank units are equipped with 14 ignitron switches instead of gap switches to attenuate the discharge noise level, an acceleration current of 400 kA with a short rise time of 9 {mu}s is controlled within a jitter of much less than 1 {mu}s. The resulting CT velocity and mass density satisfy the requirements for CT penetration into the tokamak plasma core at a toroidal field of 1 T. This CT injection system is thus suitable for CT injection in a middle-sized tokamak plasma such as the JFT-2M tokamak.

Compact toroid injection system for JFT-2M

International Nuclear Information System (INIS)

Fukumoto, N.; Ogawa, H.; Nagata, M.; Uyama, T.; Shibata, T.; Kashiwa, Y.; Suzuki, S.; Kusama, Y.

2006-01-01

The compact toroid (CT) injection system for JFT-2M is composed of a CT injector, a gas delivery and vacuum system, a power supply system, and a diagnostics system. In particular, the power supply system delivers high performance for CT formation and acceleration. The CT formation capacitor bank unit achieved a formation current of 350 kA with a rise time less than 10 μs. Although the CT acceleration bank units are equipped with 14 ignitron switches instead of gap switches to attenuate the discharge noise level, an acceleration current of 400 kA with a short rise time of 9 μs is controlled within a jitter of much less than 1 μs. The resulting CT velocity and mass density satisfy the requirements for CT penetration into the tokamak plasma core at a toroidal field of 1 T. This CT injection system is thus suitable for CT injection in a middle-sized tokamak plasma such as the JFT-2M tokamak

Robotics in the nuclear environment-inspection and repairs inside the primary coolant system

International Nuclear Information System (INIS)

Guillet, J.; Marcel Tortolano

2005-01-01

The increase in the lifetime of the power plants and the ageing of materials require the intervention inside the components to carry out controls and possibly repairs in the event of discovered defects. Within this framework, EDF is investigating the feasibility of robotized repairs of the components and pipes of the main primary coolant system of a nuclear power plant. For several years, EDF R and D has engaged projects whose subject of study is the possibility of repairing components such as the main vessel; the pressurizer or the primary coolant pipes with the help of robots and dedicated tools. INTERVENTIONS INSIDE PRIMARY COOLANT PIPES: Studies undertaken by EDF highlighted that certain zones, particularly in pipe connections, can be affected by thermal fatigue which causes crackling defects or crackings. In anticipation of this phenomenon which would affect primary pipes and to avoid their replacements, EDF R and D has been studying the feasibility of examining and repairing these zones using robots. Robotized repair consists in introducing into the pipe while passing by the vessel, a 6 degrees of freedom manipulator mounted on a mobile carrier. This robot implements and carries out the trajectories of the different processes of repair: - Precise localization of the defects, - Elimination (possibly sampling) of the defects by machining, - Control that the defects were eliminated, - Weld metal buildup if the repair cavity is too deep, - Grinding followed by a new control of the surface. These studies and tests were conducted in the laboratory of EDF R and D in Chatou. The sequence of operations included machining by grinding and milling, profilometric control, dye penetrant testing, TIG welding and ultrasonic examinations. The results of the tests, executed on full scale models of components, are satisfactory and show the advantages of robotics compared with classical methods. ROBOTIZED INTERVENTIONS IN THE REACTOR VESSEL: Another difficult issue is the

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.

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.)

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

Coolant channel module CCM

International Nuclear Information System (INIS)

Hoeld, Alois

2007-01-01

. This package can be adopted as a general element in the simulation of thermal-hydraulic situations of complex systems consisting of a number of special channels. Such systems can represent different types of steam generators, 3D nuclear reactor cores with special attention to the calculation of the mass flow distribution into different parallel channels after non-symmetric perturbations, each of them distinguished by their key numbers. The resulting set of equations can be combined with other ODE-s and constitutive equations from additional parts of such a comprehensive model. The complete system of equations can then (outside of the CCM) be solved by applying appropriate integration routines. Verification and validation test runs over a wide application range have yielded very satisfactory results demonstrating therefore in a convincing way the quality of the CCM. This approach offers an alternative to the currently dominant 'Separate-Phase Models' where each phase within a coolant channel is treated separately. The advantages and disadvantages of these two approaches are discussed in this paper

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