ORTURB, HTGR Steam Turbine Dynamic for FSV Reactor

International Nuclear Information System (INIS)

Conklin, J.C.

2001-01-01

1 - Description of program or function: ORTURB was written specifically to calculate the dynamic behavior of the Fort St. Vrain (FSV) High- Temperature Gas-Cooled Reactor (HTGR) steam turbines. The program is divided into three main parts: the driver subroutine; turbine subroutines to calculate the pressure-flow balance of the high-, intermediate-, and low-pressure turbines; and feedwater heater subroutines. 2 - Method of solution: The program uses a relationship derived for ideal gas flow in an iterative fashion that minimizes computational time to determine the pressure and flow in the FSV steam turbines as a function of plant transient operating conditions. An important computer modeling characteristic, unique to FSV, is that the high-pressure turbine exhaust steam is used to drive the reactor core coolant circulators prior to entering the reheater. A feedwater heater dynamic simulation model utilizing seven state variables for each of the five heaters is included in the ORTURB computer simulation of the regenerative Rankine cycle steam turbines. The seven temperature differential equations are solved at each time- step using a matrix exponential method. 3 - Restrictions on the complexity of the problem: The turbine shaft is assumed to rotate at a constant (rated) speed of 3600 rpm. Energy and mass storage of steam in the high-, intermediate-, and low-pressure turbines is assumed to be negligible. These limitations exclude the use of ORTURB during a turbine transient such as startup from zero power or very low turbine flows

Effects of aging and service wear on main steam isolation valves and valve operators

International Nuclear Information System (INIS)

Clark, R.L.

1996-03-01

In recent years main steam isolation valve (MSIV operating problems have resulted in significant operational transients (e.g., spurious reactor trips, steam generator dry out, excessive valve seat leakage), increased cost, and decreased plant availability. A key ingredient to an engineering-oriented reliability improvement effort is a thorough understanding of relevant historical experience. A detailed review of historical failure data available through the Institute of Nuclear Power Operation's Nuclear Plant Reliability Data System has been conducted for several types of MSIVs and valve operators for both boiling-water reactors and pressurized-water reactors. The focus of this review is on MSIV failures modes, actuator failure modes, consequences of failure on plant operations, method of failure detection, and major stressors affecting both valves and valve operators

A study on ex-vessel steam explosion for a flooded reactor cavity of reactor scale - 15216

International Nuclear Information System (INIS)

Song, S.; Yoon, E.; Kim, Y.; Cho, Y.

2015-01-01

A steam explosion can occur when a molten corium is mixed with a coolant, more volatile liquid. In severe accidents, corium can come into contact with coolant either when it flows to the bottom of the reactor vessel and encounters the reactor coolant, or when it breaches the reactor vessel and flows into the reactor containment. A steam explosion could then threaten the containment structures, such as the reactor vessel or the concrete walls/penetrations of the containment building. This study is to understand the shortcomings of the existing analysis code (TEXAS-V) and to estimate the steam explosion loads on reactor scale and assess the effect of variables, then we compared results and physical phenomena. Sensitivity study of major parameters for initial condition is performed. Variables related to melt corium such as corium temperature, falling velocity and diameter of melt are more important to the ex-vessel steam explosion load and the steam explosion loads are proportional to these variables related to melt corium. Coolant temperature on reactor cavity has a specific area to increase the steam explosion loads. These results will be used to evaluate the steam explosion loads using ROAAM (Risk Oriented Accident Analysis Methodology) and to develop the evaluation methodology of ex-vessel steam explosion. (authors)

Simulation of a main steam line break with steam generator tube rupture using trace

Energy Technology Data Exchange (ETDEWEB)

Gallardo, S.; Querol, A.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Camino de Vera s/n, 46022, Valencia (Spain)

2012-07-01

A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation of the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)

Fast reactor steam generators with sodium on the tube side. Design and operational parameters

International Nuclear Information System (INIS)

1994-01-01

A comparison of design and operational characteristics as well as analysis of experience gained during the long terms operation of the Micro Module Inverse Steam Generator and Module Inverse Steam Generator at BOR 60 reactor are main aims of this technical report. 20 refs, 47 figs, 14 tabs

Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

Directory of Open Access Journals (Sweden)

Fic Adam

2015-03-01

Full Text Available Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle, which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle. The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

Frequency and distribution of leakages in steam generators of gas-cooled reactors

International Nuclear Information System (INIS)

Bongratz, R.; Breitbach, G.; Wolters, J.

1988-01-01

In gas cooled reactors with graphitic primary circuit structures - such as HTR, AGR or Magnox - the water ingress is an event of great safety concern. Water or steam entering the primary circuit react with the hot graphite and carbon-oxide and hydrogen are produced. As the most important initiating event a leak in a steam generator must be taken into account. From the safety point of view as well as for availability reasons it is necessary to construct reliable boilers. Thus the occurrence of a boiler leak should be a rare event. In the context of a probabilistic safety study for an HTR-Project much effort was invested to get information about the frequency and the size distribution of tube failures in steam generators of gas cooled reactors. The main data base was the boiler tube failure statistics of United Kingdom gas cooled reactors. The data were selected and applied to a modern HTR steam generator design. A review of the data showed that the failure frequency is not connected with the load level (pressures, temperatures) or with the geometric size of the heating surface of the boiler. Design, construction, fabrication, examination and operation conditions have the greatest influence an the failure frequency but they are practically not to be quantified. The typical leak develops from smallest size. By erosion effects of the entering water or steam it is enlarged to perhaps some mm 2 , then usually it is detected by moisture monitors. Sudden tube breaks were not reported in the investigated period. As a rule boiler leaks in gas cooled reactors are much more, rare then leaks in steam generators of light water reactors and fossil fired boilers. (author)

Composite electric generator equipped with steam generator for heating reactor coolant

International Nuclear Information System (INIS)

Watabe, Masaharu; Soman, Yoshindo; Kawanishi, Kohei; Ota, Masato.

1997-01-01

The present invention concerns a composite electric generator having coolants, as a heating source, of a PWR type reactor or a thermonuclear reactor. An electric generator driving gas turbine is disposed, and a superheater using a high temperature exhaust gas of the gas turbine as a heating source is disposed, and main steams are superheated by the superheater to elevate the temperature at the inlet of the turbine. This can increase the electric generation capacity as well as increase the electric generation efficiency. In addition, since the humidity in the vicinity of the exit of the steam turbine is reduced, occurrence of loss and erosion can be suppressed. When cooling water of the thermonuclear reactor is used, the electric power generated by the electric generator driven by the gas turbine can be used upon start of the thermonuclear reactor, and it is not necessary to dispose a large scaled special power source in the vicinity, which is efficient. (N.H.)

Superheated steam annealing of pressurized water reactor vessel

International Nuclear Information System (INIS)

Porowski, J.S.

1993-01-01

Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 deg. F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 deg. F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors. Dry thermal annealing consists of heating portions of the reactor vessel at a specific temperature for a given period of time using a high temperature heat source. The use of spent fuel assemblies, induction heating and resistance heating elements as well as the circulation of heated fluid were investigated as potential candidate methods. To date the use of resistance heating elements which are lowered into a dry empty reactor was considered to be the preferred method. In-depth research in the United States and practical applications of such a method in Russia have confirmed feasibility of the method. The method of using circulating superheated steam to anneal the vessel at 850 deg. F without complete removal of the reactor internals is described herein. After removing the reactor head and fuel, the core barrel along with the upper and lower core in PWRs is lifted to open an annular space between the reactor shell flange and the core barrel flange. The thermal shield can remain

Studying the processes of sodium-water interaction in the BOR-60 reactor micromodule steam generator

International Nuclear Information System (INIS)

Tsykanov, V.A.; Antipin, G.K.; Borisov, V.V.

1981-01-01

Main results of experimental studies of emergency regimes of micromodule steam generator (MSG) at small and big leaks of water into sodium, realized using the 30 MW MSG, operating in the BOR-o0 reactor, are considered. The aims of the study are as follows: the modelling of macroleak in ''Nadja'' steam generator for the BN-350 reactor; testing the conceptions of alarm signalling and MSG protection; testing under real conditions of new perspective systems of leak detection; gaining the experimence and development of the ways to eliminate the consequences of accident caused by big water leak into sodium; accumulation of knowledge on restoration of MSG operating ability after accident; experimental test of calculational techniques for big leak accidents to use them in future for calculational studies of similar situations at other reactors equipped with sodium-water steam generators; refinement of characteristics of hydrodynamic and thermal effects interaction zone for big leak in real circuit during the plant operation. A series of experiments with the imitation of water leak into sodium by means of argon and steam supply through injection devices, located before the steam superheater module of one of the sections and between evaporator module of the same section, is conducted. The range of steam flow rate is 0.02-0.45 g/s. Duration of steam supply is 100-400 s. A conclusion is made that the results obtained can be used for steam generator of the BN-350 reactor [ru

Crack formation in ferritic screws of main steam isolation valves in German boiling water reactors

International Nuclear Information System (INIS)

Steinmill, H.

1992-01-01

In connection with crack formations at screws of main steam isolation valves in boiling water reactors, detected for the first time in 1988 in the Federal Republic of Germany, metallographic and fractographic investigations and coating analyses of screw surfaces and crack flanks were performed in order to find out the causes. These and other investigations of damaged screws were accompanied in the years 1989 and 1990 by autoclave tests made in several laboratories. With a view to the mechanical stress of the screws, tightening tests and stress analyses were performed by means of FEM. Repeated autoclave tests were concluded recently by the Stuttgart MPA. Although these tests are not reported here, it can be stated that the results obtained fit in with the overall framework of the results summed up in this report. With regard to the kind of sample stress and the results obtained, two cases have to be distinguished in the autoclave tests discussed in this report. (orig.) [de

Main steam system piping response under safety/relief valve opening events

International Nuclear Information System (INIS)

Swain, E.O.; Esswein, G.A.; Hwang, H.L.; Nieh, C.T.

1980-01-01

The stresses in the main steam branch pipe of a Boiling Water Reactor due to safety/relief valve blowdown has been measured from an in situ piping system test. The test results were compared with analytical results. The predicted stresses using the current state of art analytical methods used for BWR SRV discharge transient piping response loads were found to be conservative when compared to the measured stress values. 3 refs

Nuclear reactor steam depressurization valve

International Nuclear Information System (INIS)

Moore, G.L.

1991-01-01

This patent describes improvement in a nuclear reactor plant, an improved steam depressurization valve positioned intermediate along a steam discharge pipe for controlling the venting of steam pressure from the reactor through the pipe. The improvement comprises: a housing including a domed cover forming a chamber and having a partition plate dividing the chamber into a fluid pressure activation compartment and a steam flow control compartment, the valve housing being provided with an inlet connection and an outlet connection in the steam flow control compartment, and a fluid duct in communication with a source of fluid pressure for operating the valve; a valve set mounted within the fluid flow control compartment comprising a cylindrical section surrounding the inlet connection with one end adjoining the connection and having a radially projecting flange at the other end with a contoured extended valve sealing flange provided with an annular valve sealing member, and a valve cylinder traversing the partition plate and reciprocally movable within an opening in the partition plate with one terminal and extending into the fluid pressure activation compartment and the other terminal end extending into the steam flow control compartment coaxially aligned with the valve seat surrounding the inlet connection, the valve cylinder being surrounded by two bellow fluid seals and provided with guides to inhibit lateral movement, an end of the valve cylinder extending into the fluid flow control compartment having a radially projecting flange substantially conterminous with the valve seat flange and having a contoured surface facing and complimentary to the contoured valve seating surface whereby the two contoured valve surfaces can meet in matching relationship, thus providing a pressure actuated reciprocatable valve member for making closing contact with the valve seat and withdrawing therefrom for opening fluid flow through the valve

Study on steam separation in steam generators of a NPP with the WWER-440 reactors

International Nuclear Information System (INIS)

Dmitriev, A.I.; Kolzov, Yu.V.; Titov, V.F.; Dubrovin, A.V.; Ilyushin, V.F.; Volkov, A.P.

1977-01-01

The separation characteristics as well as the actual level position in steam generators with and without a submerged holy sheet have been determined at a WWER-440 reactor nuclear power plant. It has been shown, that without changing the design of steam generators their load at the WWER-440 reactor nuclear power plant can be increased by about 10%. In this case the vapour humidity does not exceed the permissible value equal to 0.25%. The submerged holy sheet considerably decreases load irregularity and swelling of the water-steam mixture layer

A reflux capsule steam generator for sodium cooled reactors

International Nuclear Information System (INIS)

Lantz, E.

Pressurized water reactor plants at numerous sites have sustained significant leakage through their steam generators. The consequent shutdowns for repairs and replacements have damaged their economics. This experience suggests that if steam generators for liquid metal fast breeder reactors (LMFBR's) continue to be built as presently designed some of them will have similar problems. Because of their larger capital investment, the consequent damage to the economics of LMFBR's could be more serious. Reflux capsules provide a way to separate sodium from water and to reduce thermal stresses in steam generators for sodium cooled reactors. Their use would also eliminate the need for a primary heat exchanger and a secondary sodium loop pump. (author)

Analysis of steam explosions in plate-type, uranium-aluminum fuel test reactors

International Nuclear Information System (INIS)

Taleyarkhan, R.P.

1989-01-01

The concern over steam explosions in nuclear reactors can be traced to prompt critical nuclear excursions in aluminum-clad/fueled test reactors, as well as to explosive events in aluminum, pulp, and paper industries. The Reactor Safety Study prompted an extensive analytical and experimental effort for over a decade. This has led to significant improvements in their understanding of the steam explosion issue for commercial light water reactors. However, little progress has been made toward applying the lessons learned from this effort to the understanding and modeling of steam explosion phenomena in aluminum-clad/fueled research and test reactors. The purposes of this paper are to (a) provide a preliminary analysis of the destructive events in test reactors, based on current understandings of steam explosions; (b) provide a proposed approach for determining the likelihood of a steam explosion event under scenarios in which molten U-Al fuel drops into a water-filled cavity; and (c) present a benchmarking study conducted to estimate peak pressure pulse magnitudes

Steam-generator-tube-rupture transients for pressurized-water reactors

International Nuclear Information System (INIS)

Dobranich, D.; Henninger, R.J.; DeMuth, N.S.

1982-01-01

Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures

Effects of Secondary Circuit Modeling on Results of Pressurized Water Reactor Main Steam Line Break Benchmark Calculations with New Coupled Code TRAB-3D/SMABRE

International Nuclear Information System (INIS)

Daavittila, Antti; Haemaelaeinen, Anitta; Kyrki-Rajamaeki, Riitta

2003-01-01

All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important

Steam turbine chemistry in light water reactor plants

International Nuclear Information System (INIS)

Svoboda, Robert; Haertel, Klaus

2008-01-01

Steam turbines in boiling water reactor (BWR) and pressurized water reactor (PWR) power plants of various manufacturers have been affected by corrosion fatigue and stress corrosion cracking. Steam chemistry has not been a prime focus for related research because the water in nuclear steam generating systems is considered to be of high purity. Steam turbine chemistry however addresses more the problems encountered in fossil fired power plants on all volatile treatment, where corrosive environments can be formed in zones where wet steam is re-evaporated and dries out, or in the phase transition zone, where superheated steam starts to condense in the low-pressure (LP) turbine. In BWR plants the situation is aggravated by the fact that no alkalizing agents are used in the cycle, thus making any anionic impurity immediately acidic. This is illustrated by case studies of pitting corrosion of a 12 % Cr steel gland seal and of flow-oriented corrosion attack on LP turbine blades in the phase transition zone. In PWR plants, volatile alkalizing agents are used that provide some buffering of acidic impurities, but they also produce anionic decomposition products. (orig.)

Modeling and simulation of an isothermal reactor for methanol steam reforming

Directory of Open Access Journals (Sweden)

Raphael Menechini Neto

2014-04-01

Full Text Available Due to growing electricity demand, cheap renewable energy sources are needed. Fuel cells are an interesting alternative for generating electricity since they use hydrogen as their main fuel and release only water and heat to the environment. Although fuel cells show great flexibility in size and operating temperature (some models even operate at low temperatures, the technology has the drawback for hydrogen transportation and storage. However, hydrogen may be produced from methanol steam reforming obtained from renewable sources such as biomass. The use of methanol as raw material in hydrogen production process by steam reforming is highly interesting owing to the fact that alcohol has the best hydrogen carbon-1 ratio (4:1 and may be processed at low temperatures and atmospheric pressures. They are features which are desirable for its use in autonomous fuel cells. Current research develops a mathematical model of an isothermal methanol steam reforming reactor and validates it against experimental data from the literature. The mathematical model was solved numerically by MATLAB® and the comparison of its predictions for different experimental conditions indicated that the developed model and the methodology for its numerical solution were adequate. Further, a preliminary analysis was undertaken on methanol steam reforming reactor project for autonomous fuel cell.

CATHENA analysis of CANDU 6 steam generators for steam main break at a remote location

Energy Technology Data Exchange (ETDEWEB)

Liu, T. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2009-07-01

CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a nonequilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. It is the primary Thermalhydraulics network analysis tool used by Atomic Energy Canada Ltd. (AECL) in the design, safety and licensing analysis of power and research reactors as well as test facilities. In the thermalhydraulic model, the liquid and vapor phases may have different pressures, velocities, and temperatures. The objective of the present paper is to present the detailed modeling of a CANDU 6 Steam Generator (SG) using the transient, thermalhydraulics network code CATHENA. The model represents the secondary side, primary side and the main steam system including the main steam line up to the assumed break location. The present model is designed such that the transient pressure drops across Tube Support Plates (TSP) could be extracted. The resistances of degraded/fouled TSPs were modeled by using the increased/reduced flow area of the TSPs. CATHENA then calculates the flow resistance in two-phase flow based on the area contraction/expansion at the TSPs. Three sets of simulations were performed; one with the degraded steam generator data provided by the utility users, and the other two with waterlanced (cleaned with high pressure water jet) TSPs. One run assumed the flow area increased by 25 percent, the other run assumed the flow area increased by 50 percent. on the hot side of the SG. No significant changes in the break discharge rates were observed between the runs. However, the steam generator downcomer flow for the waterlanced case did not reverse during the blowdown as was calculated for the degraded case. As expected, the pressure drop across the TSPs was decreased in the waterlanced cases comparing with degraded cases. The CATHENA simulation provides estimates of the velocity, density, and quality in the tube bundle as well as

Questions raised in developing fast reactor steam generator designs

Energy Technology Data Exchange (ETDEWEB)

Taylor, P A; Hayden, O

1975-07-01

The most important component in the achievement of satisfactory LMFBR reliability is the steam generator. When the failure statistics of other nuclear steam generators and the implications of a sodium water reaction are considered, there is some cause for concern. It is apparent that considerable improvement in technology is necessary and until more experience on operating plant is available a conservative design approach must be taken. Many solutions have been proposed, varying from forced circulation straight tube modular to large single vessel once through helical designs. The paper poses what are considered to be the main questions which arise when making a choice of fast reactor steam generator type and tube configuration. The aim is to promote discussion amongst the assembled experts on their relative design approaches and the importance placed upon the various factors in reaching our common goal of ensuring the success of the LMFBR in its essential role of conserving world energy resources. (author)

Questions raised in developing fast reactor steam generator designs

International Nuclear Information System (INIS)

Taylor, P.A.; Hayden, O.

1975-01-01

The most important component in the achievement of satisfactory LMFBR reliability is the steam generator. When the failure statistics of other nuclear steam generators and the implications of a sodium water reaction are considered, there is some cause for concern. It is apparent that considerable improvement in technology is necessary and until more experience on operating plant is available a conservative design approach must be taken. Many solutions have been proposed, varying from forced circulation straight tube modular to large single vessel once through helical designs. The paper poses what are considered to be the main questions which arise when making a choice of fast reactor steam generator type and tube configuration. The aim is to promote discussion amongst the assembled experts on their relative design approaches and the importance placed upon the various factors in reaching our common goal of ensuring the success of the LMFBR in its essential role of conserving world energy resources. (author)

Imitative modeling automatic system Control of steam pressure in the main steam collector with the influence on the main Servomotor steam turbine

Science.gov (United States)

Andriushin, A. V.; Zverkov, V. P.; Kuzishchin, V. F.; Ryzhkov, O. S.; Sabanin, V. R.

2017-11-01

The research and setting results of steam pressure in the main steam collector “Do itself” automatic control system (ACS) with high-speed feedback on steam pressure in the turbine regulating stage are presented. The ACS setup is performed on the simulation model of the controlled object developed for this purpose with load-dependent static and dynamic characteristics and a non-linear control algorithm with pulse control of the turbine main servomotor. A method for tuning nonlinear ACS with a numerical algorithm for multiparametric optimization and a procedure for separate dynamic adjustment of control devices in a two-loop ACS are proposed and implemented. It is shown that the nonlinear ACS adjusted with the proposed method with the regulators constant parameters ensures reliable and high-quality operation without the occurrence of oscillations in the transient processes the operating range of the turbine loads.

Clinch river breeder reactor plant steam generator water quality

Energy Technology Data Exchange (ETDEWEB)

Van Hoesen, D; Lowe, P A

1975-07-01

The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: (1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; (2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and (3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present.

Clinch river breeder reactor plant steam generator water quality

International Nuclear Information System (INIS)

Van Hoesen, D.; Lowe, P.A.

1975-01-01

The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: 1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; 2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and 3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present

Computer codes for simulation of Angra 1 reactor steam generator

International Nuclear Information System (INIS)

Pinto, A.C.

1978-01-01

A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt

Feasibility and application on steam injector for next-generation reactor

International Nuclear Information System (INIS)

Narabayashi, Tadashi; Ishiyama, Takenori; Miyano, Hiroshi; Nei, Hiromichi; Shioiri, Akio

1991-01-01

A feasibility study has been conducted on steam injector for a next generation reactor. The steam injector is a simple, compact passive device for water injection, such as Passive Core Injection System (PCIS) of Passive Containment Cooling System (PCCS), because of easy start-up without an AC power. An analysis model for a steam injector characteristics has been developed, and investigated with a visualized fundamental test for a two-stage Steam Injector System (SIS) for PCIS and a one-stage low pressure SIS for PCCS. The test results showed good agreement with the analysis results. The analysis and the test results showed the SIS could work over a very wide range of the steam pressure, and is applicable for PCIS or PCCS in the next generation reactors. (author)

AREVA main steam line break fully coupled methodology based on CATHARE-ARTEMIS - 15496

International Nuclear Information System (INIS)

Denis, L.; Jasserand, L.; Tomatis, D.; Segond, M.; Royere, C.; Sauvage, J.Y.

2015-01-01

The CATHARE code developed since 1979 by AREVA, CEA, EDF and IRSN is one of the major thermal-hydraulic system codes worldwide. In order to have at disposal realistic methodologies based on CATHARE for the whole transient and accident analysis in Chapter 15 of Safety Reports, a coupling with the code ARTEMIS was developed. ARTEMIS is the core code in AREVA's new reactor simulator system ARCADIA, using COBRA-FLX to model the thermal-hydraulics in the core. The Fully Coupled Methodology was adapted to the CATHARE-ARTEMIS coupling to perform Main Steam Line Break studies. This methodology, originally applied to the MANTA-SMART-FLICA coupling, is dedicated to Main Steam Line Break transients at zero power. The aim of this paper is to present the coupling between CATHARE and ARTEMIS and the application of the Fully Coupled Methodology in a different code environment. (authors)

Prototype fast breeder reactor main options

International Nuclear Information System (INIS)

Bhoje, S.B.; Chellapandi, P.

1996-01-01

Fast reactor programme gets importance in the Indian energy market because of continuous growing demand of electricity and resources limited to only coal and FBR. India started its fast reactor programme with the construction of 40 MWt Fast Breeder Test Reactor (FBTR). The reactor attained its first criticality in October 1985. The reactor power will be raised to 40 MWt in near future. As a logical follow-up of FBTR, it was decided to build a prototype fast breeder reactor, PFBR. Considering significant effects of capital cost and construction period on economy, systematic efforts are made to reduce the same. The number of primary and secondary sodium loops and components have been reduced. Sodium coolant, pool type concept, oxide fuel, 20% CW D9, SS 316 LN and modified 9Cr-1Mo steel (T91) materials have been selected for PFBR. Based on the operating experience, the integrity of the high temperature components including fuel and cost optimization aspects, the plant temperatures are recommended. Steam temperature of 763 K at 16.6 MPa and a single TG of 500 MWe gross output have been decided. PFBR will be located at Kalpakkam site on the coast of Bay of Bengal. The plant life is designed for 30 y and 75% load factor. In this paper the justifications for the main options chosen are given in brief. (author). 2 figs, 2 tabs

Steam up over reactor policy

International Nuclear Information System (INIS)

Kovan, D.

1976-01-01

Britain is once more assessing its nuclear power programme in the light of recent forecasts that there is unlikely to be any growth in the demand for electricity for many years to come. This means that the extra costs of launching a commercially unproven reactor, the Steam Generating Heavy Water Reactor (SGHWR), will be an even greater burden than previously expected, because they would be spread over fewer reactors. Sir John Hill's reported assessment concludes that the present strategy would be the most expensive way of developing Britain's nuclear power programme; and under the circumstances, may not be the best option. The SGHWR programme will certainly be more expensive than either relaunching a programme of advanced gas-cooled reactors (AGRs), or building American designed pressurised water reactors (PWRs). Recent developments of the AGR and PWR's and their advantages in the present position are outlined. (U.K.)

Performance Evaluation of a Printed Circuit Steam Generator for Integral Reactors: A Feasibility Test

Energy Technology Data Exchange (ETDEWEB)

Han, Hun Sik; Kang, Han-Ok; Yoon, Juhyeon; Kim, Young In; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of); Seo, Jang-won; Choi, Brain [Alfa Laval Korea Ltd., Daejeon (Korea, Republic of)

2015-05-15

SMART (System-integrated Modular Advanced ReacTor) is a small-sized integral type pressurized water reactor. It adopts advanced design features such as structural safety improvement, system simplification, and component modularization to achieve highly enhanced safety and improved economics. The design issues related to further safety enhancement and cost reduction have received significant attention to increase its competitiveness in the global small reactor market. For the cost reduction, it is important to design the reactor vessel as small as possible. Thus, it is necessary to reduce the volume of main components such as a steam generator. Its manufacturing processes of the chemical etching and diffusion bonding provide high effectiveness, high compactness, and inherent structural safety under high temperatures and high pressures. Thus, it is expected to be an alternative to the conventional shell and tube type steam generator in SMART. In this paper, simple thermal-hydraulic performance measurement of a small-scale printed circuit steam generator (PCSG) is conducted to investigate the feasibility of applying it to SMART. The simple thermal-hydraulic performance of the PCSG has been experimentally evaluated. A small-scale PCHE is employed to investigate the feasibility of operating it as a steam generator. The performance assessment reveals that the PCSG stably produces superheated steam, and the increased degree of superheat is obtained at lower water flow rate. However, the flow instability is increased with the decrease of the water flow rate. Thus, it is required to apply the orifice design into the cold side plate to suppress the density-wave oscillations. The pressure drops and heat transfer rates increase with the water flow rate.

Lungmen ABWR containment analyses during short-term main steam line break LOCA using GOTHIC

International Nuclear Information System (INIS)

Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che

2012-01-01

Highlights: ► The Lungmen ABWR containment responses due to the main steam line break are analyzed. ► In the Lungmen FSAR, the peak drywell temperature is greater than the designed value. ► GOTHIC is used to calculate the containment responses in this study. ► With more realistic conditions, the drywell temperature can be reasonably suppressed. - Abstract: Lungmen Nuclear Power Plant in Taiwan is a GE-designed twin-unit Advanced Boiling Water Reactor (ABWR) plant with rated thermal power of 3926 MWt. Both units are currently under construction. In the Lungmen Final Safety Analysis Report (FSAR) section 6.2, the calculated peak drywell temperature during the short-term Main Steam Line Break (MSLB) event is 176.3 °C, which is greater than the designed temperature of 171.1 °C. It resulted in a controversial issue in the FSAR review process conducted by the Atomic Energy Council in Taiwan. The purpose of this study is to independently investigate the Lungmen ABWR containment pressure and temperature responses to the MSLB using the GOTHIC program. Blowdown conditions are either calculated by using a simplified reactor vessel volume in GOTHIC model, or provided by the RELAP5 transient analysis. The blowdown flow rate from the steam header side is calculated with a more reasonable pressure loss coefficient of the open main steam isolation valves, and the peak drywell temperature is then reduced. By using the RELAP5 blowdown data, the peak drywell temperature can be further reduced because of the initial liquid entrainment in the blowdown flow. The drywell space is either treated as a single volume, or separated into a upper drywell and a lower drywell to reflect the real configuration of the Lungmen containment. It is also found that a single drywell volume may not present the overheating of the upper drywell. With more realistic approaches and assumptions, the drywell temperature can be reasonably below the design value and the Lungmen containment integrity

Development of a nuclear steam generator system for gas-cooled reactors for application in oil sands extraction

International Nuclear Information System (INIS)

Smith, J.; Hart, R.; Lazic, L.

2009-01-01

Canada has vast energy reserves in the Oil Sands regions of Alberta and Saskatchewan. Present extraction technologies, such as strip mining, where oil deposits are close to the surface, and Steam Assisted Gravity Drainage (SAGD) technologies for deeper deposits consume significant amounts of energy to produce the bitumen and upgraded synthetic crude oil. Studies have been performed to assess the feasibility of using nuclear reactors as primary energy sources to produce, in particular the steam required for the SAGD deeper deposit extraction process. Presently available reactors fall short of meeting the requirements, in two areas: the steam produced in a 'standard' reactor is too low in pressure and temperature for the SAGD process. Requirements can be for steam as high as 12MPa pressure with superheat; and, 'standard' reactors are too large in total output. Ideally, reactors of output in the range of 400 to 500 MWth, in modules are better suited to Oil Sands applications. The above two requirements can be met using gas-cooled reactors. Generally, newer generation gas-cooled reactors have been designed for power generation, using Brayton Cycle gas turbines run directly from the heated reactor coolant (helium). Where secondary steam is required, heat recovery steam generators have been used. In this paper, a steam generating system is described which uses the high temperature helium from the reactor directly for steam generation purposes, with sufficient quantities of steam produced to allow for SAGD steam injection, power generation using a steam turbine-generator, and with potential secondary energy supply for other purposes such as hydrogen production for upgrading, and environmental remediation processes. It is assumed that the reactors will be in one central location, run by a utility type organization, providing process steam and electricity to surrounding Oil Sands projects, so steam produced is at very high pressure (12 MPa), with superheat, in order to

Wet steam turbines for CANDU-Reactors

International Nuclear Information System (INIS)

Westmacott, C.H.L.

1977-01-01

The technical characteristics of 4 wet steam turbine aggregates used in the Pickering nuclear power station are reported on along with operational experience. So far, the general experience was positive. Furthermore, plans are mentioned to use this type of turbines in other CANDU reactors. (UA) [de

Steam leak detection in advance reactors via acoustics method

International Nuclear Information System (INIS)

Singh, Raj Kumar; Rao, A. Rama

2011-01-01

Highlights: → Steam leak detection system is developed to detect any leak inside the reactor vault. → The technique uses leak noise frequency spectrum for leak detection. → Testing of system and method to locate the leak is also developed and discussed in present paper. - Abstract: Prediction of LOCA (loss of coolant activity) plays very important role in safety of nuclear reactor. Coolant is responsible for heat transfer from fuel bundles. Loss of coolant is an accidental situation which requires immediate shut down of reactor. Fall in system pressure during LOCA is the trip parameter used for initiating automatic reactor shut down. However, in primary heat transport system operating in two phase regimes, detection of small break LOCA is not simple. Due to very slow leak rates, time for the fall of pressure is significantly slow. From reactor safety point of view, it is extremely important to find reliable and effective alternative for detecting slow pressure drop in case of small break LOCA. One such technique is the acoustic signal caused by LOCA in small breaks. In boiling water reactors whose primary heat transport is to be driven by natural circulation, small break LOCA detection is important. For prompt action on post small break LOCA, steam leak detection system is developed to detect any leak inside the reactor vault. The detection technique is reliable and plays a very important role in ensuring safety of the reactor. Methodology developed for steam leak detection is discussed in present paper. The methods to locate the leak is also developed and discussed in present paper which is based on analysis of the signal.

Technical and economic studies of small reactors for supply of electricity and steam

International Nuclear Information System (INIS)

Spiewak, I.; Klepper, O.H.; Fuller, L.C.

1977-01-01

Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out of several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 313 MW(t) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000 MW(t), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500 MW(t) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating or desalination

Technical and economic studies of small reactors for supply of electricity and steam

International Nuclear Information System (INIS)

Spiewak, I.; Klepper, O.H.; Fuller, L.C.

1977-02-01

Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out for several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 365 MW(th) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000 MW(th), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500 MW(th) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule, and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating, or desalination

Technical and economic studies of small reactors for supply of electricity and steam

International Nuclear Information System (INIS)

Spiewak, I.; Klepper, O.H.; Fuller, L.C.

1977-01-01

Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out of several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 313MW(th) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000MW(th), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500MW(th) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating or desalination. (author)

Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

Science.gov (United States)

McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

1994-01-01

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

Gas-cooled nuclear reactor

International Nuclear Information System (INIS)

1974-01-01

The invention aims at simplying gas-cooled nuclear reactors. For the cooling gas, the reactor is provided with a main circulation system comprising one or several energy conversion main groups such as gas turbines, and an auxiliary circulation system comprising at least one steam-generating boiler heated by the gas after its passage through the reactor core and adapted to feed a steam turbine with motive steam. The invention can be applied to reactors the main groups of which are direct-cycle gas turbines [fr

Development and application of an entrainment model for the PWR U-tube steam generators for main steam line break analysis

International Nuclear Information System (INIS)

Song, Dong-Soo; Park, Young-Chan

2004-01-01

The purpose of this paper is to present the analyses that were performed to develop and use an entrainment model for pressurized water reactor U-tube steam generators (SG) for main steam line break (MSLB) analyses. The entrainment model was developed using the RETRAN-3D computer program, and the model was benchmarked against experimental data of moisture carryover during a simulated MSLB accident. The application methodology was also developed to incorporate into the MSLB mass and energy release calculations for Kori Unit 1. This methodology utilizes LOFTRAN and RETRAN-3D codes in an iterative sequence of cases in which the LOFTRAN nuclear steam supply system model provides boundary conditions for the RETRAN-3D broken loop steam generator model, and the RETRAN-3D model provides the entrainment data that is input back into the LOFTRAN model. FORTRAN programs were developed to facilitate the sequencing of these iterative calculations. As a result of applying the entrainment model to Kori Unit 1, the temperature calculated inside Containment during MSLB accident using the CONTEMP-LT computer program decreased by about 25degC. Consequently this entrainment model provides a significant benefit by decreasing the temperature envelop for environment qualification as well as decreasing the peak Containment pressure. (author)

Development of phenomena identification and ranking table for APR1400 main steam line break

International Nuclear Information System (INIS)

Song, J. H.; Chung, B. D.; Jeong, J. J.

2003-01-01

A Phenomena Identification and Ranking Table (PIRT) was developed for the Main Steam Line Break (MSLB) event of an APR-1400 (Advanced Power Reactor-1400). A team of experts from research institutes, industries, and regulatory bodies participated in the development. The selected event was a double-ended steam line break at full power with the reactor coolant pump running. The panel selected the fuel performance as the primary safety criterion for ranking. The plant design data, the results of APR-1400 safety analysis, and the results of additional best estimate analysis by MARS2.1 were utilized. Three phases of pre-trip, rapid cool-down, and safety injection phase were identified. Then, the ranking of a system, components, phenomenon/process based on the relative importance to the primary evaluation criterion were followed for each time phase. Finally, the knowledge-level for each important process in the component was ranked in terms of the existing knowledge. The highly ranked phenomena identified for APR-1400 MSLB are tube wall heat transfer at the steam generator shell, void distribution at the steam generator shell, liquid entrainment in the separators, mixture level in the separators, boron mixing in the upper down comer, boron transport and thermal mixing in the lower plenum, stored energy release in the upper head, and flow to and/from the upper head. The PIRT will be used as a guide in planning cost effective experimental programs and code development efforts, especially for the quantification of the process and/or phenomena, which have a high importance but low knowledge level

Emergency systems and protection equipment of modular steam generators for fast reactors

International Nuclear Information System (INIS)

Matal, O.

The requirements are discussed for accident protection of modular steam generators for fast reactors. Accident protection is assessed for a modular through-flow steam generator and for a natural circulation modular steam generator. Benefits and constraints are shown and possible improvements are outlined for accident protection of liquid sodium fired modular steam generators. (Kr)

Detection of steam generator tube leaks in pressurized water reactors

International Nuclear Information System (INIS)

Roach, W.H.

1985-01-01

This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in PWRs, formed the basis of study for the last year of the project. Four tasks are addressed in this study of the detection of steam tube leaks. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above. (4) Assessing the need for diagnostic data processing and display. Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation. (orig./HP)

Analysis of the Nonlinear Density Wave Two-Phase Instability in a Steam Generator of 600MWe Liquid Metal Reactor

International Nuclear Information System (INIS)

Choi, Seok Ki; Kim, Seong O

2011-01-01

A 600 MWe demonstration reactor being developed at KAERI employs a once-through helically coiled steam generator. The helically coiled steam generator is compact and is efficient for heat transfer, however, it may suffer from the two-phase instability. It is well known that the density wave instability is the main source of instability among various types of instabilities in a helically coiled S/G in a LMR. In the present study a simple method for analysis of the density wave two phase instability in a liquid metal reactor S/G is proposed and the method is applied to the analysis of density wave instability in a S/G of 600MWe liquid metal reactor

The Steam Generating Heavy Water Reactor

International Nuclear Information System (INIS)

Middleton, J.E.

1975-01-01

An account is given of the SGHWR, the prototype of which was built by the United Kingdom Atomic Energy Authority at Winfrith, under the following headings: Introduction; origin of the SGHWR concept; conceptual design (choice of reactor type, steam cycle, reactor coolant system, nuclear behaviour, fuel design, core design, and protective, auxiliary and containment systems); operation and control (integrity of core cooling, reactivity control, power trimming, long term reactivity control, xenon override, load following, power shaping, spatial stability control, void coefficient); protective systems (breached coolant circuit trip, intact coolant circuits trip, power set-back trip); dynamic characteristics; reactor control; station control (decoupled control system, coupled control system, rate of response); Winfrith prototype (design and safety philosophy, conceptual features and parameters, reactor coolant system, protective systems, emergency core cooling, core structure, fuel design, vented containment). (U.K.)

Steam explosion - physical foundations and relation to nuclear reactor safety

International Nuclear Information System (INIS)

Schumann, U.

1982-08-01

'Steam explosion' means the sudden evaporation of a fluid by heat exchange with a hotter material. Other terms are 'vapour explosion', 'thermal explosion', and 'energetic fuel-coolant interaction (FCI)'. In such an event a large fraction of the thermal energy initially stored in the hot material may possibly be converted into mechanical work. For pressurized water reactors one discusses (e.g. in risk analysis studies) a core melt-down accident during which molten fuel comes into contact with water. In the analysis of the consequences one has to investigate steam explosions. In this report an overview over the state of the knowledge is given. The overview is based on an extensive literature review. The objective of the report is to provide the basic knowledge which is required for understanding of the most important theories on the process of steam explosions. Following topics are treated: overview on steam explosion incidents, work potential, spontaneous nucleation, concept of detonation, results of some typical experiments, hydrodynamic fragmentation of drops, bubbles and jets, coarse mixtures, film-boiling, scenario of a core melt-down accident with possible steam-explosion in a pressurized water reactor. (orig.) [de

Study of condensation heat transfer following a main steam line break inside containment

Energy Technology Data Exchange (ETDEWEB)

Cho, J.H.; Elia, F.A. Jr.; Lischer, D.J. [Stone & Webster Engineering Corporation, Boston, MA (United States)

1995-09-01

An alternative model for calculating condensation heat transfer following a main stream line break (MSLB) accident is proposed. The proposed model predictions and the current regulatory model predictions are compared to the results of the Carolinas Virginia Tube Reactor (CVTR) test. The very conservative results predicted by the current regulatory model result from: (1) low estimate of the condensation heat transfer coefficient by the Uchida correlation and (2) neglecting the convective contribution to the overall heat transfer. Neglecting the convection overestimates the mass of steam being condensed and does not permit the calculation of additional convective heat transfer resulting from superheated conditions. In this study, the Uchida correlation is used, but correction factors for the effects of convection an superheat are derived. The proposed model uses heat and mass transfer analogy methods to estimate to convective fraction of the total heat transfer and bases the steam removal rate on the condensation heat transfer portion only. The results predicted by the proposed model are shown to be conservative and more accurate than those predicted by the current regulatory model when compared with the results of the CVTR test. Results for typical pressurized water reactors indicate that the proposed model provides a basis for lowering the equipment qualification temperature envelope, particularly at later times following the accident.

An integrated approach to steam condensation studies inside reactor containments: A review

Energy Technology Data Exchange (ETDEWEB)

Yadav, Mahesh Kumar [Department of Mechanical Engineering, Indian Institute of Technology Kanpur, 208016 (India); Khandekar, Sameer, E-mail: samkhan@iitk.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Kanpur, 208016 (India); Sharma, Pavan K. [Reactor Safety Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

2016-04-15

Occurrence of severe accidents, such as the Fukushima incident in 2011, is unlikely with a probability of 10{sup −5} per reactor per year. However, such kinds of accidents have serious consequences on both, short term as well as on long term public health, environment and energy policy and security. They also adversely affect the progress of nuclear power industry. Thus, despite such a low probability of occurrence, a need arises to review the safety standards of nuclear power plants, especially in the light of the Fukushima accident. Apart from other systems, a review of thermal-hydraulics and safety system for the reactor containment is vital, as it is the last barrier to radioactive leakage. Main threats to the containment integrity include over-pressurization, not only due to steam alone, but its coupling with the possibility of local hydrogen combustion, depending on the local mixture composition of steam-air-hydrogens. It must be emphasized that steam condensation rate affects the local mixture composition and presence of hydrogen significantly deteriorates the condensation rate. This intrinsic coupling needs to be understood. In this paper, steam condensation and related issues, including basics of condensation, modeling approaches, parameters affecting condensation and experiments performed (in both separate effect and integral test facilities) are critically reviewed, in the light of coupled issues of hydrogen transport and combustion. Such studies are necessary for correlation development and/or to find out the local distribution of steam-hydrogen-air mixture within the containment to locate the possible hydrogen combustion location(s) and hence, deployment of active/passive safety systems. In addition, it is important that future studies, both experimental and numerical modeling, focus on the coupled nature of the problem in a comprehensive manner for ensuring long term safety.

The modification of main steam safety valves in Qinshan phase Ⅱ expansion project

International Nuclear Information System (INIS)

Chen Haiqiao

2012-01-01

The main steam safety valves of NPP steam system are second- class nuclear safety component. It used to limit the pressure of SG secondary side and main steam system via emitting steam into the environment. At present, the main steam safety valves have mechanical valves and assisted power valves. According to the experience of power plants at home and abroad, including Qinshan Phase Ⅱ unit 1/2 experience feedback, Qinshan Phase Ⅱ expansion project made modification on valve type, setting value and valve body. This paper introduce the characteristics of different safety valve types, the modification of main steam safety valves and the modification analysis on safety issues.security and impact on the other systems in Qinshan Phase Ⅱ expansion project. (author)

Investigation of thermodynamic cycle for generic 1200 MW{sub el} pressure channel reactor with nuclear steam superheat

Energy Technology Data Exchange (ETDEWEB)

Vincze, A.; Sidawi, K.; Abdullah, R.; Baldock, M.; Saltanov, E.; Pioro, I., E-mail: andrei.vincze@uoit.net, E-mail: khalil.sidawi@uoit.net, E-mail: rand.abdullah@uoit.net, E-mail: matthew.baldock@uoit.net, E-mail: eugene.saltanov@uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)

2014-07-01

Current Nuclear Power Plants (NPPs) play a significant role in energy production around the world. All NPPs operating today employ a Rankine steam cycle for the conversion of thermal power to electricity. This paper will examine the steam cycle arrangement an experimental pressure channel reactor using Nuclear Steam Superheat (NSS) and compare it to two advanced reactor designs, the Advanced CANDU Reactor 1000 (ACR-1000) and the Advanced Boiling Water Reactor (ABWR) designs. The thermodynamic cycle layout and thermal efficiencies of the three reactor types will be discussed. (author)

Steam generator tube failures

International Nuclear Information System (INIS)

MacDonald, P.E.; Shah, V.N.; Ward, L.W.; Ellison, P.G.

1996-04-01

A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service

Pressurized-water coolant nuclear reactor steam generator

International Nuclear Information System (INIS)

Mayer, H.; Schroder, H.J.

1975-01-01

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

Screening reactor steam/water piping systems for water hammer

International Nuclear Information System (INIS)

Griffith, P.

1997-09-01

A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greater than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made

Hydrogen production by biomass steam gasification in fluidized bed reactor with Co catalyst

International Nuclear Information System (INIS)

Kazuhiko Tasaka; Atsushi Tsutsumi; Takeshi Furusawa

2006-01-01

The catalytic performances of Co/MgO catalysts were investigated in steam gasification of cellulose and steam reforming of tar derived from cellulose gasification. For steam reforming of cellulose tar in a secondary fixed bed reactor, 12 wt.% Co/MgO catalyst attained more than 80% of tar reduction. The amount of produced H 2 and CO 2 increased with the presence of catalyst, and kept same level during 2 hr at 873 K. It is indicated that steam reforming of cellulose tar proceeds sufficiently over Co/MgO catalyst. For steam gasification of cellulose in a fluidized bed reactor, it was found that tar reduction increases with Co loading amount and 36 wt.% Co/MgO catalyst showed 84% of tar reduction. The amounts of produced gas kept for 2 hr indicating that 36 wt.% Co/MgO catalyst is stable during the reaction. It was concluded that these Co catalysts are promising systems for the steam gasification of cellulose and steam reforming of cellulose tar. (authors)

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

International Nuclear Information System (INIS)

Bennert, J.; Becher, L.

1977-01-01

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

IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Working Material

International Nuclear Information System (INIS)

2011-01-01

The IAEA, within the framework of its Nuclear Energy Department’s Technical Working Group on Fast Reactors (TWG-FR), assists Member States activities in fast reactors technology development areas by providing an umbrella for information exchange [topical Technical Meetings (TMs), Workshops and large Conferences] and collaborative R&D [Coordinated Research Projects (CRPs)]. The Technical meeting on “Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors” was held from 21 – 22 December 2011 in Vienna, addressing Member States’ expressed needs of information exchange in the field of advanced fast reactor design features, with particular attention to innovative heat exchangers and steam generators. The Objective of the TM is to provide a global forum for in-depth information exchange and discussion on the most advanced concepts of heat exchangers and steam generators for fast reactors. More specifically, the objectives are: · Review of the status of advanced fast reactor development activities with special emphasis on design and performance of heat exchangers and steam generators; · Discuss requirements for innovative heat exchangers and steam generators; · Present results of studies and conceptual designs for innovative heat exchangers and steam generators; · Provide recommendations for international collaboration under the IAEA aegis. The meeting agenda of the meeting is in Annex I

HTGR steam generator development

International Nuclear Information System (INIS)

Schuetzenduebel, W.G.; Hunt, P.S.; Weber, M.

1976-01-01

More than 40 gas-cooled reactor plants have produced in excess of 400 reactor years of operating experience which have proved a reasonably high rate of gas-cooled reactor steam generator availability. The steam generators used in these reactors include single U-tube and straight-tube steam generators as well as meander type and helically wound or involute tube steam generators. It appears that modern reactors are being equipped with helically wound steam generators of the once-through type as the end product of steam generator evolution in gas-cooled reactor plants. This paper provides a general overview of gas-cooled reactor steam generator evolution and operating experience and shows how design criteria and constraints, research and development, and experience data are factored into the design/development of modern helically wound tube steam generators for the present generation of gas-cooled reactors

Development of ferritic steels for steam generators of fast breeder reactors

International Nuclear Information System (INIS)

Nguyen-Thanh; Vigneron, G.; Vanderschaeghe, A.

1988-01-01

STEIN INDUSTRIE, a manufacturer of equipment for the conventional and nuclear power industry, has built up expertise in the use of Cr-Mo steels used at high temperatures. The main ferritic steels developed were 10 CD 9-10 (AFNOR), Z10 CDNb V 9-2 (AFNOR), X 20 Cr Mo V 12-1 (DIN) and ASTM Grade 9.1. For the fast breeder reactor system, STEIN INDUSTRIE proposes the use of these steels in the construction of steam generators. The wide programme of development undertaken by STEIN INDUSTRIE is aimed at the following main subjects: - characterization of materials - welding and bending tests - studies of special junctions. This article reports the results obtained

HTGR power plant hot reheat steam pressure control system

International Nuclear Information System (INIS)

Braytenbah, A.S.; Jaegtnes, K.O.

1975-01-01

A control system for a high temperature gas cooled reactor (HTGR) power plant is disclosed wherein such plant includes a plurality of steam generators. Dual turbine-generators are connected to the common steam headers, a high pressure element of each turbine receiving steam from the main steam header, and an intermediate-low pressure element of each turbine receiving steam from the hot reheat header. Associated with each high pressure element is a bypass line connected between the main steam header and a cold reheat header, which is commonly connected to the high pressure element exhausts. A control system governs the flow of steam through the first and second bypass lines to provide for a desired minimum steam flow through the steam generator reheater sections at times when the total steam flow through the turbines is less than such minimum, and to regulate the hot reheat header steam pressure to improve control of the auxiliary steam turbines and thereby improve control of the reactor coolant gas flow, particularly following a turbine trip. (U.S.)

Analysis of fast reactor steam generator performance

International Nuclear Information System (INIS)

Hulme, G.; Curzon, A.F.

1992-01-01

A computer model for the prediction of flow and temperature fields within a fast reactor steam generator unit is described. The model combines a commercially available computational fluid dynamics (CFD) solver (PHOENICS) with a steam-tube calculation and provides solutions for the fully coupled flow and temperature fields on both the shell side and the tube side. The model includes the inlet and outlet headers and the bottom end stagnant zone. It also accounts for the effects of support grids and edge-gaps. Two and three dimensional and transient calculations have been performed for both straight tube and J-tube units. Examples of the application of the model are presented. (7 figures) (Author)

On the heat exchange tube failures in steam generators at NPPs with WWER reactors

International Nuclear Information System (INIS)

Titov, V.F.; Banyuk, G.F.; Brykov, S.I.

1992-01-01

Data on dynamics of failed heat exchanging tube closing in steam generators of NPPs with WWER type reactors for the whole period of their operation are presented. It is shown that the main cause of the tube failures consists in their corrosion cracking under stresses. The effect of chlorine ions on tubes is intensified by the presence of porous sediments on heat exchaning surfaces in quantities exceeding 150 g/m 2

HEXTRAN-SMABRE calculation of the 6th AER Benchmark, main steam line break in a WWER-440 NPP

International Nuclear Information System (INIS)

Haemaelaeinen, A.; Kyrki-Rajamaeki, R.

2003-01-01

The sixth AER benchmark is the second AER benchmark for couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a double end break of one main steam line in a WWER-440 plant. The core is at the end of its first cycle in full power conditions. In VTT HEXTRAN2.9 is used for the core kinetics and dynamics and SMABRE4.8 as a thermal hydraulic model for the primary and secondary loop. The plant model for SMABRE consists mainly of two input models, Loviisa model and a standard WWER-440/213 plant model. The primary loop includes six separate loops, the pressure vessel is divided into six parallel channels in SMABRE and the whole core calculation is performed in the core with HEXTRAN. The horizontal steam generators are modelled with heat transfer tubes in five levels and vertically with two parts, riser and downcomer. With this kind of detailed modelling of steam generators there occurs strong flashing after break opening. As a sequence of the main steam line break at nominal power level, the reactor trip is followed quite soon. The liquid temperature continues to decrease in one core inlet sector which may lead to recriticality and neuron power increase. The situation is very sensitive to small changes in the steam generator and break flow modelling and therefore several sensitivity calculations have been done. Also two stucked control rods have been assumed. Due to boric acid concentration in the high pressure safety injection subcriticality is finally guaranteed in the transient (Authors)

Evaluation of material integrity on electricity generator water steam cycles component (Main Steam Pipe)

International Nuclear Information System (INIS)

Sudardjo; Histori; Triyadi, Ari

1998-01-01

The evaluation of material integrity on electricity generator component has been done. That component was main steam pipe of Unit II Suralaya Coal Fired Power Plant. evaluation was done by replication technique. The damage was found are two porosity's, from two point samples of six points sample population. Based on cavity evaluation in steels, which proposed by Neubauer and Wedel that porosity's still at class A damage. For class A damage, its means no remedial action would be required until next major scheduled maintenance outage. That porosity's was grouped on isolated cavities and not need ti repair that main steam pipe component less than three year after replication test

Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

Directory of Open Access Journals (Sweden)

Yan Wang

2017-01-01

Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

The effect of steam cycle conditions upon the economics and design of a sodium-cooled fast reactor

International Nuclear Information System (INIS)

Philpott, E.F.; Pounder, F.; Willby, C.R.

1978-01-01

The paper studies the effect of variation of steam and feedwater conditions upon the economics, design and layout of a sodium-cooled fast reactor. The parameters investigated are steam temperature and pressure, feedwater temperature, and boiler recirculation ratio. The paper also includes an assessment of the effects of associating the fast reactor with saturated steam cycle conditions. (author)

Calculation of steam content in a draught section of a tank-type boiling water cooled reactor

International Nuclear Information System (INIS)

Panajotov, D.P.; Gorburov, V.I.

1989-01-01

Structural and hydrodynamic features of a two-phase flow in a draught section of a tank-type boiling water cooled reactor are considered. A calculated model of the steady flow and methods for determining steam content and phase rate profiles under the maximum steam content at the section axis and at some distance from it are proposed. Steam content distribution by height quantitatively agrees with experimental data for the VK-50 reactor. Calculation technique allows one to obtain steam content and phase rate profiles at the section outlet

Analyses in support of installation of steam-dump-to-atmosphere valves at steam lines of the Dukovany NPP

International Nuclear Information System (INIS)

Kral, P.

1998-01-01

Four conservative analyses were carried out with a view to examining the cooldown capacity of the super-emergency feedwater pump (SEFWP) → steam generator (SG) → steam dump to atmosphere/main steam line (SDA/MSL) chain. This emergency cooldown capacity was investigated for a postulated accident associated with a main steam header break + main feedwater header break + closing of all main steam lines, and for an artificial accident with SCRAM + isolation of all MSLs + loss of feedwater. The RELAP5/MOD3.1 code and a detailed 3-loop input model of the Dukovany plant were employed. Conservative assumptions with respect to the initial reactor power, decay heat evolution, and other input parameters were applied. The results gave evidence that the capacity of both the 2SEFWP → 2SG → 2SDA/SG and 1SEFWP → 1SG → 1SDA/SG chains is sufficient for the decay heat to be removed from the reactor; however, a considerably long time allowing for a sufficient drop of the decay heat is necessary for a deep cooldown of the primary circuit. For the event encompassing main steam header break + main feedwater header break with isolation of all MSLs and with cooling by 2SEFWPs, a time-consuming calculation gave evidence of the feasibility of passing to the water-water regime and primary system cooldown to below 93 deg C in the hot legs

Steam conversion of liquefied petroleum gas and methane in microchannel reactor

Science.gov (United States)

Dimov, S. V.; Gasenko, O. A.; Fokin, M. I.; Kuznetsov, V. V.

2018-03-01

This study presents experimental results of steam conversion of liquefied petroleum gas and methane in annular catalytic reactor - heat exchanger. The steam reforming was done on the Rh/Al2O3 nanocatalyst with the heat applied through the microchannel gap from the outer wall. Concentrations of the products of chemical reactions in the outlet gas mixture are measured at different temperatures of reactor. The range of channel wall temperatures at which the ratio of hydrogen and carbon oxide in the outlet mixture grows substantially is determined. Data on the composition of liquefied petroleum gas conversion products for the ratio S/C = 5 was received for different GHVS.

Process for superheating the steam generated by a light water nuclear reactor

International Nuclear Information System (INIS)

Vakil, H.B.; Brown, D.H.

1976-01-01

A process is submitted for superheating the pressurised steam generated in a light water nuclear reactor in which the steam is brought to 340 0 C at least. This superheated steam is used to operate a turbo-generator unit. The characteristic of the process is that an exothermal chemical reaction is used to generate the heat utilised during the superheating stage. The chemical reaction is a mechanisation, oxidation-reduction or hydrogenation reaction [fr

New safety experiments in decommissioned superheated steam reactor at Karlstein

International Nuclear Information System (INIS)

Koerting, K.

1986-01-01

This article gives a concise summary of the Status Report of the Superheated Steam Reactor Safety Program (PHDR) Project, held at KfK on Dec. 5, 1985. The results discussed dealt with fire experiments, shock tests simulating airplane crashes, temperature shocks in the reactor pressure vessel, studies of crack detection in pressure vessels and blasting experiments associated with nuclear plant decommissioning

Some aspects affecting fast reactor steam generator integrity considered from a utility viewpoint

Energy Technology Data Exchange (ETDEWEB)

Bolt, P R

1975-07-01

The important conditions affecting fast reactor steam generator integrity are discussed. In addition to the need for high integrity levels when the steam generator is first delivered to the power station site, the equally important aspect of demonstrating retention of continued high levels of integrity throughout the operating life of the station is described. The functional and related conditions that are believed important to the selection of a design type which can offer adequately high levels of integrity are given. Some of the data needs of a utility concerned with fast reactor S.G.U. design assessment are described, particular emphasis being given to areas believed to have a significant effect on steam generator reliability and integrity. (author)

Steam generator for a pressurized-water coolant nuclear reactor

International Nuclear Information System (INIS)

Schroeder, H.J.; Berger, W.

1975-01-01

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

Heat exchanging tube behaviour in steam generators of pressurized water reactors

International Nuclear Information System (INIS)

Pastor, D.; Oertel, K.

1979-01-01

Based on a comprehensive failure statistics, materials corrosion chemistry and thermohydraulics problems of the tubings of steam generators are considered. A historical review of failures in the tubings of steam generators in pressurized water reactors reflects the often successless measures by designers, manufacturers and operating organizations for preventing failures, especially with regard to materials selection and water regime. It is stated that laboratory tests could not give sufficient information about safe and stable operation of nuclear steam generators unless real constructive, hydrodynamic, thermodynamical and chemical conditions of operation had been taken into account. (author)

Reactor operation method

International Nuclear Information System (INIS)

Osumi, Katsumi; Miki, Minoru.

1979-01-01

Purpose: To prevent stress corrosion cracks by decreasing the dissolved oxygen and hydrogen peroxide concentrations in the coolants within a reactor container upon transient operation such as at the start-up or shutdown of bwr type reactors. Method: After a condensate has been evacuated, deaeration operation is conducted while opening a main steam drain line, as well as a main steam separation valve and a by-pass valve in a turbine by-pass line connecting the main steam line and the condenser without by way of a turbine, and the reactor is started-up by the extraction of control rods after the concentration of dissolved oxygen in the cooling water within a pressure vessel has been decreased below a predetermined value. Nuclear heating is started after the reactor water has been increased to about 150 0 C by pump heating after the end of the deaeration operation for preventing the concentration of hydrogen peroxide and oxygen in the reactor water from temporarily increasing immediately after the start-up. The corrosive atmosphere in the reactor vessel can thus be moderated. (Horiuchi, T.)

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.)

Steam generator of FBR type reactor

International Nuclear Information System (INIS)

Hashiguchi, Ko.

1992-01-01

Liquid metal (for example, mercury) which is scarcely reactive with metal sodium is contained and cover gases which are scarcely reactive with the liquid metal are filled in a steam generator of an FBR type reactor and it is closed. The heat of primary sodium is transferred to the liquid metal, which is not reactive with sodium, in a primary thermal conduction portion. Since the temperature of the primary thermal conduction portion is high, the density is extremely low. On the other hand, since a second thermal conduction portion is kept at a single phase and the temperature is lower compared with that of the first thermal conduction portion, the density is kept high. since the density difference and gas jetting speed generate a great circulating force to liquid metal passing the opening of a partition plate, heat can be conducted on the side of water without disposing pumps. The steam concentration in the liquid metal is low being in a single phase of steams, corrosion caused from the outside of pipes of the primary thermal conduction pipe is scarcely promoted. Even if sodium leaks should be caused, since the sodium concentration in the liquid metal is extremely low and the reactivity is low, the temperature of the liquid metal is not elevated. (N.H.)

Reducing scram frequency by modifying/eliminating steam generator low-low level reactor trip setpoint for Maanshan nuclear power plant

International Nuclear Information System (INIS)

Yuann, R.Y.; Chiang, S.C.; Hsiue, J.K.; Chen, P.C.

1987-01-01

The feasibility of modification/elimination of steam generator low-low level reactor trip setpoint is evaluated by using RETRAN-02 code for the purpose of reducing scram frequency in Maanshan 3-loop pressurized water reactor. The ANS Condition II event loss of normal feedwater and condition IV event feedwater system line break are the basis for steam generator low-low level reactor trip setpoint sensitivity analysis, including various initial reactor power levels, reactivity feedback coefficients, and system functions assumptions etc., have been performed for the two basis events with steam generator low-low level reactor trip setpoint at 0% narrow range and without this trip respectively. The feasibility of modifying/eliminating current steam generator low-low level reactor trip setpoint is then determined based on whether the analysis results meet with the ANS Condition II and IV acceptance criteria or not

Operating experience of main steam isolation valves at Fessenheim and Bugey

International Nuclear Information System (INIS)

Dredemis, G.; Fourest, B.; Giroux, C.

1985-07-01

The paper presents the experience of Hopkinson MSIVs over about 40 reactor-years (1977 to 1984) of operation at Fessenheim and Bugey units (900 MWe PWR). The various problems encountered including ageing effects on auxiliary equipments and increases in closure time are discussed. The corrective actions undertaken by the utility and the safety assessment of these events performed by the french safety authorities are also described. This study is the synthesis of an in-depth analysis of Main Steam Isolation Valves (MSIV) and their auxiliary circuits equipping the Bugey and Fessenheim 900 MWe PWR nuclear power plants. These valves are different from those installed in the other French 900 MWe PWR reactors. The evaluation of the operation of these valves was made on the basis of incidents which occured during operation of the units or during the periodic tests, as well as anomalies discovered during maintenance operations. This analysis proved that the anomalies related to the design of the valves, as well as to their manufacture and installation, had been correctly dealt with. Furthermore, it should have also revealed potential anomalies due to ageing of the equipment

Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type boiling water reactor

International Nuclear Information System (INIS)

Jain, Vikas; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

2011-01-01

Highlights: → We have highlighted the problem of drum level dynamics in a multiple loop type NC system using RELAP5 code. → The need of interconnections in steam and liquid spaces close to drum is established. → The steam space interconnections equalize pressure and liquid space interconnections equalize level. → With this scheme, the system can withstand anomalous conditions. → However, the controller is found to be inevitable for inventory balance. - Abstract: Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main heat transport system (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam-water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam-water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.

Online monitoring of steam/water chemistry of a fast breeder test reactor

International Nuclear Information System (INIS)

Subramanian, K.G.; Suriyanarayanan, A.; Thirunavukarasu, N.; Naganathan, V.R.; Panigrahi, B.S.; Jambunathan, D.

2005-01-01

Operating experience with the once-through steam generator of a fast breeder test reactor (FBTR) has shown that an efficient water chemistry control played a major role in minimizing corrosion related failures of steam generator tubes and ensuring steam generator tube integrity. In order to meet the stringent feedwater and steam quality specifications, use of fast and sensitive online monitors to detect impurity levels is highly desirable. Online monitoring techniques have helped in achieving feedwater of an exceptional degree of purity. Experience in operating the online monitors in the steam/water system of a FBTR is discussed in detail in this paper. In addition, the effect of excess hydrazine in the feedwater on the steam generator leak detection system and the need for a hydrazine online meter are also discussed. (orig.)

On the reliability of steam generator performance at nuclear power plants with WWER type reactors

International Nuclear Information System (INIS)

Styrikovich, M.A.; Margulova, T.Kh.

1974-01-01

The problem of ensuring reliable operation of steam generators in a nuclear power plant with a water-cooled, water-moderated reactor (WWER) was studied. At a nuclear power plant with a vertical steam generator (specifically, a Westinghouse product) the steam generator tubes were found to have been penetrated. Shutdown was due to corrosion disintegration of the austenitic stainless steel, type 18/8, used as pipe material for the heater surface. The corrosion was the result of the action of chlorine ions concentrated in the moisture contained in the iron oxide films deposited in low parts of the tube bundle, directly at the tube plate. Blowing through did not ensure complete removal of the film, and in some cases the construction features of the steam generator made removal of the film practically impossible. Replacement of type 18/8 stainless steel by other construction material, e.g., Inconel, did not give good results. To ensure reliable operation of vertical steam generators in domestic practice, the generators are designed without a low tube plate (a variant diagram of the vertical steam generator of such construction for the water-cooled, water-moderated reactor 1000 is presented). When low tube plates are used the film deposition is intolerable. For organization of a non-film regime a complex treatment of the feed water is used, in which the amount of complexion is calculated from the stoichmetric ratios with the composition of the feed water. It is noted that, if 100% condensate purification is used with complexon processing of the feed water to the generator, we can calculate the surface of the steam-generator heater without considering the outer placement on the tubes. In this the cost of the steam generator and all the nuclear power plants with WWER type reactors is decreased even with installation of a 100% condensate purification. It is concluded that only simultaneous solution of construction and water-regime problems will ensure relaible operation of

Steam relief valve control system for a nuclear reactor

International Nuclear Information System (INIS)

Torres, J.M.

1976-01-01

Described is a turbine follow system and method for Pressurized Water Reactors utilizing load bypass and/or atmospheric dump valves to provide a substitute load upon load rejection by bypassing excess steam to a condenser and/or to the atmosphere. The system generates a variable pressure setpoint as a function of load and applies an error signal to modulate the load bypass valves. The same signal which operates the bypass valves actuates a control rod automatic withdrawal prevent to insure against reactor overpower

Nuclear reactors with auxiliary boiler circuit

International Nuclear Information System (INIS)

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

1976-01-01

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

Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

International Nuclear Information System (INIS)

Jeong, Jae Jun; Chung, Bub Dong

2005-09-01

For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis

Turbine main engines

CERN Document Server

Main, John B; Herbert, C W; Bennett, A J S

1965-01-01

Turbine Main Engines deals with the principle of operation of turbine main engines. Topics covered include practical considerations that affect turbine design and efficiency; steam turbine rotors, blades, nozzles, and diaphragms; lubricating oil systems; and gas turbines for use with nuclear reactors. Gas turbines for naval boost propulsion, merchant ship propulsion, and naval main propulsion are also considered. This book is divided into three parts and begins with an overview of the basic mode of operation of the steam turbine engine and how it converts the pressure energy of the ingoing ste

In-service inspection of nuclear reactor vessels and steam generators. Results and evolution of the technics

International Nuclear Information System (INIS)

Rapin, Michel; Saglio, Robert.

1978-01-01

Methods and original technics have been developed by the CEA for inspection of the primary coolant circuit of PWR. Multifrequency Eddy currents for inspection of steam generators tubes gudgeons and bolts; focussed ultrasonics to test all the welds of the reactor vessel and its cover of mixed welds of tanks and steam generators, pressurizer welds and gudgeons from the inside; gamma radiography of vessel mixed welds, televisual examination of the stainless steel lining of the reactor vessel and its cover. Use of these technics is made with specific automatic machines designed either for inspection of steam generator tubes or for complete inspection of the vessel. Several reactors were inspected with these devices [fr

Automatic systems for opening and closing reactor vessels, steam generators, and pressurizers

International Nuclear Information System (INIS)

Samblat, C.

1990-01-01

The need for shorter working assignments, reduced dose rates and less time consumption have caused Electricite de France and Framatome to automate the entire procedure of opening and closing the main components in the primary system, such as the reactor vessel, steam generator, and pressurizer. The experience accumulated by the two companies in more than 300 annual revisions of nuclear generating units worldwide has been used as a basis for automating all bolt opening and closing steps as well as cleaning processes. The machines and automatic systems currently in operation are the result of extensive studies and practical tests. (orig.) [de

Analysis of the VVER-440 reactor steam generator secondary side with the RELAP5/MOD3 code

International Nuclear Information System (INIS)

Tuunanen, J.

1993-01-01

Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used to calculate steady state flow conditions in the secondary side of a VVER-440 reactor steam generator. (orig.) (1 ref., 9 figs., 2 tabs.)

Hydrogen production by enhanced-sorption chemical looping steam reforming of glycerol in moving-bed reactors

International Nuclear Information System (INIS)

Dou, Binlin; Song, Yongchen; Wang, Chao; Chen, Haisheng; Yang, Mingjun; Xu, Yujie

2014-01-01

Highlights: • New approach on continuous high-purity H 2 produced auto-thermally with long time. • Low-cost NiO/NiAl 2 O 4 exhibited high redox performance to H 2 from glycerol. • Oxidation, steam reforming, WSG and CO 2 capture were combined into a reactor. • H 2 purity of above 90% was produced without heating at 1.5–3.0 S/C and 500–600 °C. • Sorbent regeneration and catalyst oxidization achieved simultaneously in a reactor. - Abstract: The continuous high-purity hydrogen production by the enhanced-sorption chemical looping steam reforming of glycerol based on redox reactions integrated with in situ CO 2 removal has been experimentally studied. The process was carried out by a flow of catalyst and sorbent mixture using two moving-bed reactors. Various unit operations including oxidation, steam reforming, water gas shrift reaction and CO 2 removal were combined into a single reactor for hydrogen production in an overall economic and efficient process. The low-cost NiO/NiAl 2 O 4 catalyst efficiently converted glycerol and steam to H 2 by redox reactions and the CO 2 produced in the process was simultaneously removed by CaO sorbent. The best results with an enriched hydrogen product of above 90% in auto-thermal operation for reforming reactor were achieved at initial temperatures of 500–600 °C and ratios of steam to carbon (S/C) of 1.5–3.0. The results indicated also that not all of NiO in the catalyst can be reduced to Ni by the reaction with glycerol, and the reduced Ni can be oxidized to NiO by air at 900 °C. The catalyst oxidization and sorbent regeneration were achieved under the same conditions in air reactor

Steam generator materials performance in high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Chafey, J.E.; Roberts, D.I.

1980-11-01

This paper reviews the materials technology aspects of steam generators for HTGRs which feature a graphite-moderated, uranium-thorium, all-ceramic core and utilizes high-pressure helium as the primary coolant. The steam generators are exposed to gas-side temperatures approaching 760 0 C and produce superheated steam at 538 0 C and 16.5 MPa (2400 psi). The prototype Peach Bottom I 40-MW(e) HTGR was operated for 1349 EFPD over 7 years. Examination after decommissioning of the U-tube steam generators and other components showed the steam generators to be in very satisfactory condition. The 330-MW(e) Fort St. Vrain HTGR, now in the final stages of startup, has achieved 70% power and generated more than 1.5 x 10 6 MWh of electricity. The steam generators in this reactor are once-through units of helical configuration, requiring a number of new materials factors including creep-fatigue and water chemistry control. Current designs of larger HTGRs also feature steam generators of helical once-through design. Materials issues that are important in these designs include detailed consideration of time-dependent behavior of both base metals and welds, as required by current American Society of Mechanical Engineers (ASME) Code rules, evaluation of bimetallic weld behavior, evaluation of the properties of large forgings, etc

A dense Pd/Ag membrane reactor for methanol steam reforming: Experimental study

NARCIS (Netherlands)

Basile, A.; Gallucci, F.; Paturzo, L.

2005-01-01

This paper focuses on an experimental study of the methanol steam reforming (MSR) reaction. A dense Pd/Ag membrane reactor (MR) has been used, and its behaviour has been compared to the performance of a traditional reactor (TR) packed with the same catalyst type and amount. The parameters

Quad Cities Unit 2 Main Steam Line Acoustic Source Identification and Load Reduction

International Nuclear Information System (INIS)

DeBoo, Guy; Ramsden, Kevin; Gesior, Roman

2006-01-01

The Quad Cities Units 1 and 2 have a history of steam line vibration issues. The implementation of an Extended Power Up-rate resulted in significant increases in steam line vibration as well as acoustic loading of the steam dryers, which led to equipment failures and fatigue cracking of the dryers. This paper discusses the results of extensive data collection on the Quad Cities Unit 2 replacement dryer and the Main Steam Lines. This data was taken with the intent of identifying acoustic sources in the steam system. Review of the data confirmed that vortex shedding coupled column resonance in the relief and safety valve stub pipes were the principal sources of large magnitude acoustic loads in the main steam system. Modifications were developed in sub-scale testing to alter the acoustic properties of the valve standpipes and add acoustic damping to the system. The modifications developed and installed consisted of acoustic side branches that were attached to the Electromatic Relief Valve (ERV) and Main Steam Safety Valve (MSSV) attachment pipes. Subsequent post-modification testing was performed in plant to confirm the effectiveness of the modifications. The modifications have been demonstrated to reduce vibration loads at full Extended Power Up-rate (EPU) conditions to levels below those at Original Licensed Thermal Power (OLTP). (authors)

Optimization of the steam generator project of a gas cooled nuclear reactor

International Nuclear Information System (INIS)

Sakai, Massao

1978-01-01

The present work is concerned with the modeling of the primary and secondary circuits of a gas cooled nuclear reactor in order to obtain the relation between the parameters of the two cycles and the steam generator performance. The procedure allows the optimization of the steam generator, through the maximization of the plant net power, and the application of the optimal control theory of dynamic systems. The heat balances for the primary and secondary circuits are carried out simultaneously with the optimized - design parameters of the steam generator, obtained using an iterative technique. (author)

Reactor physics studies in the steam flooded GCFR-Phase II critical assembly

International Nuclear Information System (INIS)

Bhattacharyya, S.K.

1978-08-01

A possible accident scenario in a Gas-Cooled Fast Reactor (GCFR) is the leakage of secondary steam into the core. Considerable analytical effort has gone into the study of the effects of such an accidental steam entry. The work described represents the first full scale experimental study of the steam-entry phenomenon in GCFRs. The reference GCFR model used for the study was the benchmark GCFR Phase II assembly, and polyethylene foam was used to provide a very homogeneous steam simulation. The reactivity worth of steam entry was measured for three different steam densities. In addition, a set of integral physics parameters were measured in the largest steam density (0.008 g/cm 3 ) configuration. The corresponding parameters were also measured in dry reference GCFR critical assembly for comparison. The experiments were analyzed using ENDF/B-IV data and two-dimensional diffusion theory methods. As in earlier GCFR critical experiments analysis, the Benoist method was used to treat the problem of neutron streaming

Steam content of the two-phase flow in the Vk-50 boiling water cooled reactor draught section

International Nuclear Information System (INIS)

Fedulin, V.N.; Shmelev, V.E.; Solodkij, V.A.; Bartolomej, G.G.

1983-01-01

Results are presented of experimental investigation of the two-phase steam-water coolant flow hydrodynamics within the VK-50 reactor draught section. On the basis of the analysis of the obtained data a two-phase coolant flow model in a large diameter channel is proposed. It is shown that the steam-content distribution in the volume of the draught section has a pronounced non-equilibrium character manifested in the steam migration from the periphery to the central region. A minimum value of the steam content at the periphery is attained at the 0.7-1.0 m height; it is followed by a partial steam content levelling over the section. However the total steam content levelling over the cross section of the draught section does not take place. The steam distribution in the water layer over the draught section (overflow zone) is also nonuniform over the reactor section. The non-uniform steam distribution enchances with reduction nn pressure

Research of impact of kind resuperheat and structure of system regenerative feed water to thermodynamic efficiency of cycle with steam-coolant reactor

Directory of Open Access Journals (Sweden)

Maykova Svetlana

2017-01-01

Full Text Available The first key problems of modern nuclear reactors are inability of closed nuclear cycle, problems with spent nuclear fuel, poor effectiveness of nuclear fuel and heat-exchange equipment usage. Dealing with problems consists in usage of fast-neutron reactors with steam coolant. Scientific men analyzed neutron-physical processes in steam-cooled fast reactor and consulted that creation of the reactor is viable. In consequence of low steam activation a single-loop steam cycle may be create. The cycle is easy and fool-proof. Core thermomechanical equipment has mastered and has relatively low metal content. Results of calculation are showing that nuclear unit with steam-coolant fast neutron reactor is more efficient than widely used unit with reactor VVER. Usage of simple scheme with four regenerative feedwater heaters the absolute efficiency ratio is more than 43%.

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

International Nuclear Information System (INIS)

Zhao Xiaoyu

1996-01-01

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

Steam generator tube failures: experience with water-cooled nuclear power reactors during 1976

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1978-02-01

A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author)

Steam generator tubing development for commercial fast breeder reactors

International Nuclear Information System (INIS)

Sessions, C.E.; Uber, C.F.

1981-01-01

The development work to design, manufacture, and evaluate pre-stressed double-wall 2/one quarter/ Cr-1 Mo steel tubing for commercial fast breeder reactor steam generator application is discussed. The Westinghouse plan for qualifying tubing vendors to produce this tubing is described. The results achieved to date show that a long length pre-stressed double-wall tube is both feasible and commercially available. The evaluation included structural analysis and experimental measurement of the pre-stress within tubes, as well as dimensional, metallurgical, and interface wear tests of tube samples produced. This work is summarized and found to meet the steam generator design requirements. 10 refs

Methane-steam reforming by molten salt - membrane reactor using concentrated solar thermal energy

International Nuclear Information System (INIS)

Watanuki, K.; Nakajima, H.; Hasegawa, N.; Kaneko, H.; Tamaura, Y.

2006-01-01

By utilization of concentrated solar thermal energy for steam reforming of natural gas, which is an endothermic reaction, the chemical energy of natural gas can be up-graded. The chemical system for steam reforming of natural gas with concentrated solar thermal energy was studied to produce hydrogen by using the thermal storage with molten salt and the membrane reactor. The original steam reforming module with hydrogen permeable palladium membrane was developed and fabricated. Steam reforming of methane proceeded with the original module with palladium membrane below the decomposition temperature of molten salt (around 870 K). (authors)

Detection of steam leaks into sodium in fast reactor steam generators by acoustic techniques - An overview of Indian programme

International Nuclear Information System (INIS)

Prabhakar, R.; Vyjayanthi, R.K.; Kale, R.D.

1990-01-01

Realising the potential of acoustic leak detection technique, an experimental programme was initiated a few years back at Indira Gandhi Centre for Atomic Research (IGCAR) to develop this technique. The first phase of this programme consists of experiments to measure background noise characteristics on the steam generator modules of the 40 MW (thermal) Fast Breeder Test Reactor (FBTR) at Kalpakkam and experiments to establish leak noise characteristics with the help of a leak simulation set up. By subjecting the measured data from these experiments to signal analysis techniques, a criterion for acoustic leak detection for FBTR steam generator will be evolved. Second phase of this programme will be devoted to developing an acoustic leak detection system suitable for installation in the 500 MWe Prototype Fast Breeder Reactor (PFBR). This paper discusses the first phase of the experimental programme, results obtained from measurements carried out on FBTR steam generators and results obtained from leak simulation experiments. Acoustic leak detection system being considered for PFBR is also briefly described. 4 refs, 8 figs, 1 tab

Steam water cycle chemistry of liquid metal cooled innovative nuclear power reactors

International Nuclear Information System (INIS)

Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

2012-09-01

selection of chemistry controls is vital for NPPs with liquid metal cooled reactors. This paper highlights principles and approaches to chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors. The recommendations on how to arrange chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors are based taking into account: - the experience with operation of fossil power industry; - secondary side water chemistry of lead-bismuth eutectics cooled nuclear reactors at submarines; - steam/water cycles of NPPs with sodium cooled fast breeders BN-350 and BN-600; - secondary water chemistry at conventional NPPs with WER, RBMK and some other reactors. (authors)

RELAP5 analysis of PKL, main steam line break test

Energy Technology Data Exchange (ETDEWEB)

Jonnet, J.R.; Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; With, A. de; Wakker, P.H.

2013-12-15

Highlights: • RELAP5/MOD 3.2 code validation is performed by analyzing a main steam line break test in the PKL large scale test facility. • The RELAP5 model reproduces well the important events of the PKL test. • RELAP5 transient results show noticeable sensitivity to small differences in the initial conditions. • Accurate prediction of the coolant temperature is essential for the assessment of potential core re-criticality. - Abstract: PKL is a large scale test facility of the primary system owned by AREVA NP GmbH. It is used for extensive experimental investigations to study the integral behavior of Pressurized Water Reactor (PWR) plants under accident conditions. Since 2001, the test program is a part of an international cooperation project (SETH, followed by PKL1 and PKL2) set up by the OECD. The aim of the present work was to perform a short validation study of the thermo-hydraulics code RELAP5. A model of the PKL test facility has been developed, tested and applied to one of the experiments performed at the PKL. The chosen experiment was the test G3.1. In that experiment, a main steam line break occurs, causing a rapid depressurization of the affected steam generator. This leads to an increase of the heat transfer from the primary to the secondary side and thereby to a fast cool-down transient on the primary side. The main objective of this analysis was the qualification of the RELAP5 code results against heat transfer from the primary to the secondary side in both affected and intact loops, and temperatures in the primary system. The calculation results have been compared to the experimental results. It was concluded that the most important events during the test are reproduced relatively well by the model. The calculated coolant temperature in the core is higher than in the experiment. The minimum temperature is about 5% higher than measured. The secondary pressures in SG-1, 3, and 4 is in very good agreement with the experimental value, but in the

Development of LBB Piping Evaluation Diagram for APR 1000 Main Steam Line Piping

International Nuclear Information System (INIS)

Yang, J. S.; Jeong, I. L.; Park, C. Y.; Bai, S. Y.

2010-01-01

This paper presents the piping evaluation diagram (PED) to assess the applicability of Leak-Before- Break(LBB) for APR 1000 main steam line piping. LBB-PED of APR 1000 main steam line piping is independent of its piping geometry and has a function of the loads applied in piping system. Also, in order to evaluate LBB applicability during construction process with only the comparative evaluation of material properties between actually used and expected, the expected changes of material properties are considered in the LBB-PED. The LBB-PED, therefore, can be used for quick LBB evaluation of APR 1000 main steam line piping of both design and construction

Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-07-01

The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics.

Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

International Nuclear Information System (INIS)

1988-01-01

The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics

Improvements to PWR type reactors

International Nuclear Information System (INIS)

Ailloud, Jean; Monteil, Marcel.

1978-01-01

Improvements to pressurized water nuclear reactors are described, where the core coolant, called primary fluid, flows under the effect of a circulating pump in a primary loop between a steam generator and a pressure vessel containing the reactor core. The steam generator includes a bundle of tubes through which flows the primary fluid which exchanges calories with a secondary fluid, generally water, entering the generator as a liquid and issuing from it as steam. After expansion in turbines and recovery in a condenser, this steam is returned to the inside of the generator. Each primary fluid circulating pump is powered by a back-pressure turbine located in parallel with the high pressure section of the main turbine and hence fed with steam taken directly from the steam generator or the main steam pipe outside it [fr

Evaluation of acoustic resonance at branch section in main steam line. Part 1. Effects of steam wetness on acoustic resonance

International Nuclear Information System (INIS)

Uchiyama, Yuta; Morita, Ryo

2011-01-01

The power uprating of the nuclear power plant (NPP) is conducted in United States, EU countries and so on, and also is planned in Japan. However, the degradation phenomena such as flow-induced vibration and wall thinning may increase or expose in the power uprate condition. In U.S. NPP, the dryer had been damaged by high cycle fatigue due to acoustic-induced vibration under a 17% extended power uprating (EPU) condition. This is caused by acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSL). Increased velocity by uprating excites the pressure fluctuations and makes large amplitude resonance. To evaluate the acoustic resonance at the stub pipes of SRVs in actual BWR, it is necessary to clarify the acoustic characteristics in steam flow. Although there are several previous studies about acoustic resonance, most of them are not steam flow but air flow. Therefore in this study, to investigate the acoustic characteristics in steam flow, we conducted steam flow experiments in each dry and wet steam conditions, and also nearly saturated condition. We measured pressure fluctuation at the top of the single stub pipe and in main steam piping. As a result, acoustic resonance in dry steam flow could be evaluated as same as that in air flow. It is clarified that resonance amplitude of fluctuating pressure at the top of the stub pipe in wet steam was reduced to one-tenth compared with that in dry. (author)

Materials Options of Steam Generator for Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Fu Xiaogang; Long Bin; Han Liqing; Qin Bo; Zhang Jinquan; Wang Shuxing

2013-01-01

Overview of the material options of steam generator for sodium-cooled fast reactors, the method to calculate the service life, the thinning of wall thickness and the sodium corrosion rate, the degradation of mechanical properties (thermal aging and decarburization) and the calculation results of theoretical models

Practical use of valve seating machine with remote control system for main steam isolation valve at N.P.S

International Nuclear Information System (INIS)

Ito, Sadao; Noda, Hiroshi; Sadamura, Morito; Utsunomiya, Yasushi.

1975-01-01

The main steam isolation valves in BWR power stations are installed at the boundary of reactor containment vessels, and 2 valves in each main steam system total 8 valves in a plant. They are pneumatically operated Y type globe valves for preventing the release of radioactive substances in the atmosphere in case of the breaking of main steam pipes and also preventing the loss of coolant in case of the breaking of recirculating equipments. Therefore careful leak test, inspection, and seat-fitting are carried out to the valves at each regular maintenance. The manual maintenance work is difficult because of narrow space and the reduction of exposure, and the seat-fitting work requires the skill of high degree, therefore Okano Valve Manufacturing Co. and Tokyo Electric Power Co. jointly started the research and development of an automatic valve seating machine, and successfully put it to practical use in Fukushima No.1 Nuclear Power Station in Nov. 1974. First, the problems in the manual seat-fitting work were investigated, and the means to mechanically solve them were materialized with a prototype machine. After its mock-up test, an actual machine was designed and manufactured. The test result showed remarkable reduction of exposure and labor-saving, and the leak evaluation was sufficiently below the allowable value. (Kako, I.)

Code Assessment of SPACE 2.19 using LSTF 10% Main Steam-Line-Break Test

Energy Technology Data Exchange (ETDEWEB)

Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Hydro and Nuclear Power Co. through collaborative works with other Korean nuclear industries and research institutes. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run SBSL- 01 for a 10% main steam line break transient in a pressurized water reactor. The LSTF 10% main steam line break test were simulated using the SPACE 2.19 for code V and V work. The overall comparisons between the SPACE 2.19 code prediction and the LSTF Test Run SB-SL-01 experimental data are reasonably satisfactory. The comparisons were conducted in terms of the variations of mass flow rate, void fraction, pressure, collapsed liquid level, temperature, and system flow rate for the transient. In addition, the input model was modified for simulation accuracy of PZR pressure based on the calculated results. The correction of PORV setpoint affects to simulate the PORV open and close phenomena similarly with experiments. From the modification, the computed results show a reasonable agreement with experimental data in overall transient time.

Pd-Ag membrane reactor for steam reforming reactions: a comparison between different fuels

NARCIS (Netherlands)

Gallucci, F.; Basile, A.

2008-01-01

The simulation of a dense Pd-based membrane reactor for carrying out the methane, the methanol and the ethanol steam reforming (SR) reactions for pure hydrogen production is performed. The same simulation is also performed in a traditional reactor. This modelling work shows that the use of membrane

Methanol steam-reforming in a catalytic fixed bed reactor

Energy Technology Data Exchange (ETDEWEB)

Duesterwald, H G; Hoehlein, B; Kraut, H; Meusinger, J; Peters, R [Research Centre Juelich (KFA) (Germany). Inst. of Energy Process Engineering; Stimming, U [Technische Univ. Muenchen, Garching (Germany). Inst. fuer Festkoerperphysik und Techn. Phys.

1997-12-01

Designing an appropriate methanol steam reformer requires detailed knowledge about the processes within such a reactor. Thus, the axial temperature and concentration gradients and catalyst ageing were investigated. It was found that for a fresh catalyst load, the catalyst located in the reactor entrance was most active during the experiment. The activity of this part of the catalyst bed decreased after some time of operation due to ageing. With further operation, the most active zone moved through the catalyst bed. From the results concerning hydrogen production and catalyst degradation, the necessary amount of catalyst for a mobile PEMFC-system can be estimated. (orig.)

Development project HTR-electricity-generating plant, concept design of an advanced high-temperature reactor steam cycle plant with spherical fuel elements (HTR-K)

International Nuclear Information System (INIS)

1978-07-01

The report gives a survey of the principal work which was necessary to define the design criteria, to determine the main design data, and to design the principal reactor components for a large steam cycle plant. It is the objective of the development project to establish a concept design of an edvanced steam cycle plant with a pebble bed reactor to permit a comparison with the direct-cycle-plant and to reach a decision on the concept of a future high-temperature nuclear power plant. It is tried to establish a largerly uniform basic concept of the nuclear heat-generating systems for the electricity-generating and the process heat plant. (orig.) [de

Ethanol steam reforming kinetics of a Pd-Ag membrane reactor

NARCIS (Netherlands)

Tosti, S.; Basile, A.; Borelli, R.; Borgognoni, F.; Castelli, S.; Fabbricino, M.; Gallucci, F.; Licusati, C.

2009-01-01

The ethanol steam reforming reaction carried out in a Pd-based tubular membrane reactor has been modelled via a finite element code. The model considers the membrane tube divided into finite volume elements where the mass balances for both lumen and shell sides are carried out accordingly to the

Model for transient simulation in a PWR steam circuit

International Nuclear Information System (INIS)

Mello, L.A. de.

1982-11-01

A computer code (SURF) was developed and used to simulate pressure losses along the tubes of the main steam circuit of a PWR nuclear power plant, and the steam flow through relief and safety valves when pressure reactors its thresholds values. A thermodynamic model of turbines (high and low pressure), and its associated components are simulated too. The SURF computer code was coupled to the GEVAP computer code, complementing the simulation of a PWR nuclear power plant main steam circuit. (Author) [pt

A GASFLOW analysis of a steam explosion accident in a typical light-water reactor confinement building

International Nuclear Information System (INIS)

Travis, J.R.; Wilson, T.L.; Spore, J.W.; Lam, K.L.; Rao, D.V.

1994-01-01

Steam over-pressurization resulting from ex-vessel steam explosion (fuel-coolant interaction) may pose a serious challenge to the integrity of a typical light-water reactor confinement building. If the steam generation rate exceeds the removal capacity of the Airborne Activity Confinement System, confinement overpressurization occurs. Thus, there is a large potential for an uncontrolled and unfiltered release of fission products from the confinement atmosphere to the environment at the time of the steam explosion. The GASFLOW computer code was used to analyze the effects of a hypothetical steam explosion and the transport of steam and hydrogen throughout a typical light-water reactor confinement building. The effects of rapid pressurization and the resulting forces on the internal structures and the heat exchanger service bay hatch covers were calculated. Pressurization of the ventilation system and the potential damage to the ventilation fans and high-efficiency particulate air filters were assessed. Because of buoyancy forces and the calculated confinement velocity field, the hydrogen diffuses and mixes in the confinement atmosphere but tends to be transported to its upper region. (author). 2 refs., 14 figs

A GASFLOW analysis of a steam explosion accident in a typical light-water reactor confinement building

International Nuclear Information System (INIS)

Travis, J.R.; Wilson, T.L.; Spore, J.W.; Lam, K.L.; Rao, D.V.

1994-01-01

Steam over-pressurization resulting from ex-vessel steam explosion (fuel-coolant interaction) may pose a serious challenge to the integrity of a typical light-water reactor confinement building. If the steam generation rate exceeds the removal capacity of the Airborne Activity Confinement System, confinement over pressurization occurs. Thus, there is a large potential for an uncontrolled and unfiltered release of fission products from the confinement atmosphere to the environment at the time of the steam explosion. The GASFLOW computer code was used to analyze the effects of a hypothetical steam explosion and the transport of steam and hydrogen throughout a typical light-water reactor confinement building. The effects of rapid pressurization and the resulting forces on the internal structures and the heat exchanger service bay hatch covers were calculated. Pressurization of the ventilation system and the potential damage to the ventilation fans and high-efficiency particulate air filters were assessed. Because of buoyancy forces and the calculated confinement velocity field, the hydrogen diffuses and mixes in the confinement atmosphere but tends to be transported to its upper region

Steam drums

International Nuclear Information System (INIS)

Crowder, R.

1978-01-01

Steam drums are described that are suitable for use in steam generating heavy water reactor power stations. They receive a steam/water mixture via riser headers from the reactor core and provide by means of separators and driers steam with typically 0.5% moisture content for driving turbines. The drums are constructed as prestressed concrete pressure vessels in which the failure of one or a few of the prestressing elements does not significantly affect the overall strength of the structure. The concrete also acts as a radiation shield. (U.K.)

High-temperature gas-cooled reactor steam cycle/cogeneration application study update

International Nuclear Information System (INIS)

1981-09-01

Since publication of a report on the application of a High Temperature Gas-Cooled Reactor Steam Cycle/Cogeneration (HTGR-SC/C) plant in December of 1980, progress has continued on application related activities. In particular, a reference plant and an application identification effort has been performed, a variable cogeneration cycle balance-of-plant design was developed and an updated economic analysis was prepared. A reference HTGR-SC/C plant size of 2240 MW(t) was selected, primarily on the basis of 2240 MW(t) being in the mid-range of anticipated application needs and the availability of the design data from the 2240 MW(t) Steam Cycle/Electric generation plant design. A variable cogeneration cycle plant design was developed having the capability of operating at a range of process steam loads between the reference design load (full cogeneration) and the no process steam load condition

Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jae Jun; Chung, Bub Dong

2005-09-15

For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

Acoustic noises of the BOR-60 reactor steam generators when simulating leaks with argon and steam

International Nuclear Information System (INIS)

Sokolov, V.M.; Golushko, V.V.; Afanas'ev, V.A.; Grebenkin, Yu.P.; Muralev, A.B.

1985-01-01

Background acoustic noises of stea generators in different operational regimes and noises of argon and steam small leads (about 0.1 g/s) are studied to determine the possibility of designing the acoustic system for leak detection in sodium-water steamgenerators. Investigations are carried out at the 30 MW micromodule steam generator being in operation at the BOR-60 reactor as well as at the 20 MW tank type steam generator. Immersed ransduceres made of lithium niobate 6 mm in-diameter and waveguide transducers made of a stainless steel in the form of rods 10 mm in-diameter and 500 mm long are used as acoustic monitors. It is shown that the leak noise is more wide-band than the background noise of the steam generator and both high and low frequencies appear in the spectrum. The use of monitors of different types results in similar conslusions inrelation to the character of background noises and leak signals (spectral density, signal to-noise ratio) in the ase of similar bandroidths of the transduceres. A conclusion is made that the change of operational regimes leads to changes of background noise level, which can be close to the reaction of

Steam generator

International Nuclear Information System (INIS)

Fenet, J.-C.

1980-01-01

Steam generator particularly intended for use in the coolant system of a pressurized water reactor for vaporizing a secondary liquid, generally water, by the primary cooling liquid of the reactor and comprising special arrangements for drying the steam before it leaves the generator [fr

Membrane steam reforming of natural gas for hydrogen production by utilization of medium temperature nuclear reactor

International Nuclear Information System (INIS)

Djati Hoesen Salimy

2010-01-01

The assessment of steam reforming process with membrane reactor for hydrogen production by utilizing of medium temperature nuclear reactor has been carried out. Difference with the conventional process of natural gas steam reforming that operates at high temperature (800-1000°C), the process with membrane reactor operates at lower temperature (~500°C). This condition is possible because the use of perm-selective membrane that separate product simultaneously in reactor, drive the optimum conversion at the lower temperature. Besides that, membrane reactor also acts the role of separation unit, so the plant will be more compact. From the point of nuclear heat utilization, the low temperature of process opens the chance of medium temperature nuclear reactor utilization as heat source. Couple the medium temperature nuclear reactor with the process give the advantage from the point of saving fossil fuel that give direct implication of decreasing green house gas emission. (author)

Steam generators: critical components in nuclear steam supply systems

Energy Technology Data Exchange (ETDEWEB)

Stevens-Guille, P D

1974-02-28

Steam generators are critical components in power reactors. Even small internal leaks result in costly shutdowns for repair. Surveys show that leaks have affected one half of all water-cooled reactors in the world with steam generators. CANDU reactors have demonstrated the highest reliability. However, AECL is actively evolving new technology in design, manufacture, inspection and operation to maintain reliability. (auth)

Steam drum level control studies of a natural circulation multi loop reactor

Energy Technology Data Exchange (ETDEWEB)

Kumar, Rajesh; Contractor, A.D.; Srivastava, Abhishek; Lele, H.G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Safety Div.; Vaze, K.K. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Design and Development Group

2013-12-15

The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory. (orig.)

Intelligent information data base of flow boiling characteristics in once-through steam generator for integrated type marine water reactor

International Nuclear Information System (INIS)

Inasaka, Fujio; Nariai, Hideki

1998-01-01

Valuable experimental knowledge with flow boiling characteristics of the helical-coil type once-through steam generator was converted into an intelligent information data base program. The program was created as a windows application using the Visual Basic. Main functions of the program are as follows: (1) steady state flow boiling analysis of any helical-coil type once-through steam generator, (2) analysis and comparison with the experimental data, (3) reference and graph display of the steady state experimental data, (4) reference of the flow instability experimental data and display of the instability threshold correlated by each parameter, (5) summary of the experimental apparatus. (6) menu bar such as a help and print. In the steady state analysis, the region lengths of subcooled boiling, saturated boiling, and super-heating, and the temperature and pressure distributions etc. for secondary water calculated. Steady state analysis results agreed well with the experimental data, with the exception of the pressure drop at high mass velocity. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized water reactor with helical-coil type steam generator

Isolating valve, especially in main-steam pipes of power plants

International Nuclear Information System (INIS)

Karpenko, A.N.

1977-01-01

The valve for PWRs and BWRs, with diameters up to 1.25 m, for temperatures from -180 0 C to about 600 0 C and pressures up to over 50 bar, is designed for high reliability and long useful life. Two circular valve discs are moved as isolating elements in their plane across the steam direction and brought before the valve seat within a valve chamber. Shortly before reaching this final position, each disc is rotated by a small amount about its axis. Only after reaching the final position a double-wedge, further pushed forward between both discs, produces the necessary contact pressure. By revolving and frictionless closing caking together at high stresses and temperature variation is prevented and permanent tightness assured. The valve body is moved in a cylinder, cast on the valve housing, by means of a stepped piston. Its larger diameter is guided in a second cylinder flanged on above. In the cover of the second cylinder a pilot valve is mounted being controlled over 2 parallel solenoid valves by means of compressed air. In normal operation process steam from the valve chamber serves to move the stepped piston with the valve chamber. On closing of a bore, connecting both cylinder spaces, by the pilot valve the main valve is opened. If the pilot valve is opened the steam through the connecting bore is acting on both piston stages and closing the main valve. On loss of steam (pipe break) or for testing purposes one or the other cylinder space over solenoid valves is acted upon by auxiliary energy or evacuated, the main valve thus being controlled. (HP) [de

NPP Krsko Containment Response Following Main Steam Line Break

International Nuclear Information System (INIS)

Spalj, S.; Grgic, D.; Cavlina, N.

2000-01-01

This paper presents the calculation of thermohydraulic environmental parameters (pressure and temperature) inside containment of Krsko NPP after postulated Main Steam Line Break (MSLB) accident. This analysis was done as a part of the ambient parameters specification in the frame of the NPP Krsko Equipment Qualification (EQ) project. The RELAP5/mod2 computer code was used for the determination of MSLB mass and energy release and computer code GOTHIC was used to calculate pressure and temperature profiles inside NPP Krsko containment. The analysis was performed for spectrum of break sizes to account for possible steam superheating during accidents with smaller break sizes. (author)

Simulation of the OECD Main-Steam-Line-Break Benchmark Exercise 3 Using the Coupled RELAP5/PANTHER Codes

International Nuclear Information System (INIS)

Schneidesch, Christophe R.; Zhang Jinzhao

2004-01-01

The RELAP5 best-estimate thermal-hydraulic system code has been coupled with the PANTHER three-dimensional neutron kinetics code via the TALINK dynamic data exchange control and processing tool. The coupled RELAP5/PANTHER code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric pressurized water reactor (PWR) accidents with strong core-system interactions. The Organization for Economic Cooperation and Development/U.S. Nuclear Regulatory Commission PWR main-steam-line-break benchmark problem was analyzed as part of the qualification efforts to demonstrate the capability of the coupled code package of simulating such transients. This paper reports the main results of TE's contribution to the benchmark Exercise 3

Steam generator materials constraints in UK design gas-cooled reactors

International Nuclear Information System (INIS)

James, D.W.

1988-01-01

A widely reported problem with Magnox-type reactors was the oxidation of carbon steel components in gas circuits and steam generators. The effects of temperature, pressure, gas composition and steel composition on oxidation kinetics have been determined, thus allowing the probabilities of failure of critical components to be predicted for a given set of operating conditions. This risk analysis, coupled with regular inspection of reactor and boiler internals, has allowed continued operation of all U.K. Magnox plant. The Advanced Gas Cooled Reactor (AGR) is a direct development of the Magnox design. The first four AGRs commenced operation in 1976, at Hinkley Point 'B' and at Hunterston 'B'. All known materials problems with the steam generators have been diagnosed and solved by the development of appropriate operational strategies, together with minor plant modifications. Materials constraints no longer impose any restrictions to full load performance from the steam generators throughout the predicted life of the plant. Problems discussed in detail are: 1. oxidation of the 9 Cr - 1 Mo superheater. 2. Stress corrosion of the austenitic superheater. 3. Creep of the transition joints between the 9 Cr - 1 Mo and austenitic sections. With the 9 Cr - 1 Mo oxidation maximum temperature restriction virtually removed and creep constraints properly quantified, boiler operation in now favourably placed. Stress corrosion research has allowed the risk of tube failure to be related to time, temperature, stress and chemistry. As a result, the rigorous 'no wetting' policy has been relaxed for the normally high quality AGR feedwater, and the superheat margin has been reduced to 23 deg. C. This has increased the size of the operating window and reduced the number of expensive, and potentially harmful, plant trips. (author)

Assessment of vibration anomalies of main steam lines at Palo Verde-3

International Nuclear Information System (INIS)

Amr, A.; Landstrom, C.; Maxwell, H.; Miller, J.S.; Lynch, J.J.

1996-01-01

Historically, flow induced vibration in piping systems that transport liquid has presented problems for plant designers. When evaluating a vibration problem, it is always important to determine the forcing frequencies from different phenomena and the natural frequencies of the system as an integral part of establishing the root cause of the problem. Since in most cases of large vibration and noise levels, the natural frequency of the system and the frequency of the flow induced vibration are very close, determining the natural frequency of the system is important. Palo Verde Unit-3 exhibited a vibration problem where identification of the root cause was difficult. A Palo Verde team was created which consisted of engineers from different on-site departments and support from consultants. The process used to determine the root cause for the vibration/noise problem on Main Steam Supply System (MSSS) steam line 2 at Palo Verde Unit 3 is discussed in this paper. Since the root cause was not readily apparent, a finite element model was constructed to determine the natural frequency of the piping system. The finite element model consisted of a portion of the main steam lines, including a sample line which traverses the main steam line

Analysis of a main steam isolation valve closure anticipated transient without scram in a boiling water reactor

International Nuclear Information System (INIS)

Liaw, T.J.; Pan, C.; Chen, G.S.

1989-01-01

Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main stream isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4

Saturated steam turbines for power reactors of WWER-type

International Nuclear Information System (INIS)

Czwiertnia, K.

1978-01-01

The publication deals with design problems of large turbines for saturated steam and with problem of output limitations of single shaft normal speed units. The possibility of unification of conventional and nuclear turbines, which creates the economic basis for production of both types of turbines by one manufacturer based on standarized elements and assemblies is underlined. As separate problems the distribution of nuclear district heating power systems are considered. The choice of heat diagram for district heating saturated steam turbines, the advantages of different diagrams and evaluaton for further development are presented. On this basis a program of unified turbines both condensing and district heating type suitable for Soviet reactors of WWER-440 and WWER-1000 type for planned development of nuclear power in Poland is proposed. (author)

Solar membrane natural gas steam-reforming process: evaluation of reactor performance

NARCIS (Netherlands)

de Falco, M.; Basile, A.; Gallucci, F.

2010-01-01

In this work, the performance of an innovative plant for efficient hydrogen production using solar energy for the process heat duty requirements has been evaluated via a detailed 2D model. The steam-reforming reactor consists of a bundle of coaxial double tubes assembled in a shell. The annular

Solar membrane natural gas steam-reforming process : evaluation of reactor performance

NARCIS (Netherlands)

Falco, de M.; Basile, A.; Gallucci, F.

2010-01-01

In this work, the performance of an innovative plant for efficient hydrogen production using solar energy for the process heat duty requirements has been evaluated via a detailed 2D model. The steam-reforming reactor consists of a bundle of coaxial double tubes assembled in a shell. The annular

The Creys Malville FBR Super Phenix steam generators

International Nuclear Information System (INIS)

Baque, P.; Zuber, T.; Saur, J.M.; Cambillard, E.

1980-08-01

After briefly recalling the French experience on sodium steam generators, the authors describe the design concepts of the Superphenix units and give their main characteristics. A short summary of the realized R and D program precedes the description of the four 750-MWt steam generators, the fabrication of which is in progress by Creusot-Loire at Chalon sur Saone (France). The studies started for the next French fast breeder reactors and their steam generators are mentioned

Simulation and analysis of main steam control system based on heat transfer calculation

Science.gov (United States)

Huang, Zhenqun; Li, Ruyan; Feng, Zhongbao; Wang, Songhan; Li, Wenbo; Cheng, Jiwei; Jin, Yingai

2018-05-01

In this paper, after thermal power plant 300MW boiler was studied, mat lab was used to write calculation program about heat transfer process between the main steam and boiler flue gas and amount of water was calculated to ensure the main steam temperature keeping in target temperature. Then heat transfer calculation program was introduced into Simulink simulation platform based on control system multiple models switching and heat transfer calculation. The results show that multiple models switching control system based on heat transfer calculation not only overcome the large inertia of main stream temperature, a large hysteresis characteristic of main stream temperature, but also adapted to the boiler load changing.

Study of ex-vessel steam explosion risk of Reactor Pit Flooding System and structural response of containment for CPR1000"+ Unit

International Nuclear Information System (INIS)

Zhang Juanhua; Chen Peng

2015-01-01

Reactor Pit Flooding System is one of the special mitigation measures for severe accident for CPR1000"+ Unit. If the In-Vessel Relocation function of Reactor Pit Flooding System is failed, there is the steam explosion risk in reactor cavity. This paper firstly adopts MC3D code to build steam explosion model in order to calculate the pressure load and impulses of steam explosion that are as the input data of containment structural response analysis. The next step is to model the containment structure and analyze the structural response by ABAQUS code. The analysis results show that the integral damage induced by steam explosion to the external containment wall is shallow, and the containment structural integrity can be maintained. The risk and damage to the containment integrity reduced by steam explosion of RPF is small, and it does not influence the design and implementation of RPF. (author)

Main trends of upgrading the 1000 MW steam turbine

International Nuclear Information System (INIS)

Drahy, J.

1990-01-01

Parameters are compared for the 1000 MW steam turbine manufactured by the Skoda Works, Czechoslovakia, and turbines in the same power range by other manufacturers, viz. ABB, Siemens/KWU, GEC and LMZ. The Skoda turbine compares well with the other turbines with respect to all design parameters, and moreover, enables the most extensive heat extraction for district heating purposes. The main trends in upgrading this turbine are outlined; in particular, they include an additional increase in the heat extraction, which is made possible by a new design of the low-pressure section or by using a ''satellite'' turbine. The studies performed also indicate that the output of the full-speed saturated steam turbine can be increased to 1300 MW. An experimental turbine representing one flow of the high-pressure part of the 1000 MW turbine is being built on the 1:1 scale. It will serve to verify the methods of calculation of the wet steam flow and to experimentally test the high-pressure part over a wide span of the parameters. (Z.M.). 1 tab., 3 figs., 7 refs

Three-Dimensional Modeling of a Steam-Line Break in a Boiling Water Reactor

International Nuclear Information System (INIS)

Tinoco, Hernan

2002-01-01

Because of weld problems, the core grids of Units 1 and 2 at the Forsmark nuclear power plant have been replaced by grids of a new design, consisting of a single machined piece without welds. The qualifying structural analysis has been carried out considering dynamic loads, which implies that even loss-of-coolant accidents have to be included. Therefore, a detailed time description of the loads acting on the different internal parts of the reactor is needed. To achieve sufficient space and time resolution, a computational fluid dynamics (CFD) analysis was considered to be a viable alternative.A CFD analysis of a steam-line break in the boiling water reactor of Unit 2 is the subject of this work. The study is based on the assumption that the timescale of the transient analysis is smaller than the relaxation time of the water-steam system. Therefore, a simulation of only the upper, steam part of the reactor with no two-phase effects (flashing) is feasible.The results obtained display a rather complex behavior of the decompression process, forcing the analysis of the pressure field to be accomplished through animation. In contrast, the computed instantaneous forces over different internal parts oscillate regularly and are approximately twice the forces estimated in the past by simpler methods, with frequencies of 30 to 40 Hz; top amplitudes of ∼1.64 MN; and relatively low damping, ∼25% after 0.5 s.According to the present results, this type of modeling is physically meaningful for simulation timescales smaller than the water-steam relaxation time, i.e., ∼0.5 s at reactor conditions. At larger times, a two-phase model is necessary to describe the decompression process since two-phase effects are dominant. The results have not yet been validated with experiments, but validation computations will be run in the future for comparison with results of the Marviken tests

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

International Nuclear Information System (INIS)

Seo, Jae Kwang; Choi, Ki Yong; Kang, Han Ok; Kim, Young In; Yoon, Ju Hyeon; Zee, Sung Qunn

2005-01-01

The design features of a once-through steam generator (OTSG) for an integral reactor are significantly different from the commercial U-tube type steam generator from several aspects such as the general arrangement, size, operation conditions, and so on. Therefore a sufficient understanding of the thermal-hydraulic characteristics of the OTSG is essential for the design of the nuclear steam supply system (NSSS) and the power conversion system (PCS). It is also necessary to develop operation procedures complying to the unique design features of the OTSG of interest. The OTSG is sized to produce a sufficiently superheated steam during a normal power operation and therefore the secondary system can be simple relative to that of the other types of steam generators. For the plant adopting the OTSG, the steam pressure in the secondary circuit (tube side of the OTSG) is controlled to be constant during a normal power operation. Constant steam pressure is realized by regulating the control valve on the main steam line dedicated for this purpose. However during a heatup operation, at which the fluid state at the exit of the OTSG is a single phase hot water or two phases, it is not proper to use the control valve on the main steam line due to a control problem at low and multi-phase flow conditions and possibly an erosion problem. For these reasons, another dedicated line called a startup cooling line is used during a heatup condition. There may be several operational conditions for the secondary fluid required to pass through during heatup operation, depending on the design of the PCS. In general, there are two conditions: One is a condition for a vacuum operation for the condenser and another is an entry condition for a steam pressure control operation for an auxiliary power system. In this study, the concept of using a simple startup cooling line with a fixed flow resistance and changing the feedwater flow for the pressure control of the PCS during a heatup period are

Reactor feedwater system

International Nuclear Information System (INIS)

Kagaya, Hiroyuki; Tominaga, Kenji.

1993-01-01

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

Device for the condensation of pressurized steam and its application to the cooling of a nuclear reactor after an incident

International Nuclear Information System (INIS)

Dagard, P.; Couturier, M.

1989-01-01

This document describes an invention which relates to a device for condensation of pressurized water which is at a pressure considerably above atmospheric pressure, such as the steam produced by the steam generator of a pressurized-water nuclear reactor during the cooling of the reactor after an incident. The purpose of the invention is therefore to propose a device for the condensation of steam which is under a pressure which is considerably higher than atmospheric pressure by cooling this circulating steam as a result of contact with a heat-exchange wall which is cooled by water; such a device should be easy to install in a nuclear power plant to ensure passive cooling of the reactor, it should have a very good efficiency because of efficient heat exchangers, and it should require only a limited amount of cooling water in the equipment itself

Auditable Safety Analysis and Final Hazard Classification for the 105-N Reactor Zone and 109-N Steam Generator Zone Facility

International Nuclear Information System (INIS)

Kloster, G.L.

1998-07-01

This document is a graded auditable safety analysis (ASA) and final hazard classification (FHC) for the Reactor/Steam Generator Zone Segment. The Reactor/Steam Generator Zone Segment, part of the N Reactor Complex, that is also known as the Reactor Building and Steam Generator Cells. The installation of the modifications described within to support surveillance and maintenance activities are to be completed by July 1, 1999. The surveillance and maintenance activities addressed within are assumed to continue for the next 15- 20 years, until the initiation of facility D ampersand D (i.e., Interim Safe Storage). The graded ASA in this document is in accordance with EDPI-4.30-01, Rev. 1, Safety Analysis Documentation, (BHI-DE-1) and is consistent with guidance provided by the U.S. Department of Energy. This ASA describes the hazards within the facility and evaluates the adequacy of the measures taken to reduce, control, or mitigate the identified hazards. This document also serves as the FHC for the Reactor/Steam Generator Zone Segment. This FHC is developed through the use of bounding accident analyses that envelope the potential exposures to personnel

Development of a coupled reactor with a catalytic combustor and steam reformer for a 5 kW solid oxide fuel cell system

International Nuclear Information System (INIS)

Kang, Sanggyu; Lee, Kanghun; Yu, Sangseok; Lee, Sang Min; Ahn, Kook-Young

2014-01-01

Highlights: • Proposes the scale-up strategy to develop a large-scale coupled reactor. • Investigation of performance of steam reformer coupled with catalytic combustor. • Experimental parameters are inlet temp., air excess ratio, SCR, fuel utilization. • Evaluation of the heat transfer distribution along the gas flow direction. • The mean value of methane conversion rate is approximately 93.4%. - Abstract: The methane (CH 4 ) conversion rate of a steam reformer can be increased by thermal integration with a catalytic combustor, called a coupled reactor. In the present study, a 5 kW coupled reactor has been developed based on a 1 kW coupled reactor in previous work. The geometric parameters of the space velocity, diameter and length of the coupled reactor selected from the 1 kW coupled reactor are tuned and applied to the design of the 5 kW coupled reactor. To confirm the scale-up strategy, the performance of 5 kW coupled reactor is experimentally investigated with variations of operating parameters such as the fuel utilization in the solid oxide fuel cell (SOFC) stack, the inlet temperature of the catalytic combustor, the excess air ratio of the catalytic combustor, and the steam to carbon ratio (SCR) in the steam reformer. The temperature distributions of coupled reactors are measured along the gas flow direction. The gas composition at the steam reformer outlet is measured to find the CH 4 conversion rate of the coupled reactor. The maximum value of the CH 4 conversion rate is approximately 93.4%, which means the proposed scale-up strategy can be utilized to develop a large-scale coupled reactor

Westinghouse Small Modular Reactor nuclear steam supply system design

Energy Technology Data Exchange (ETDEWEB)

Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

Reactor abnormality diagnosis device and its method

International Nuclear Information System (INIS)

Honma, Hitoshi; Hirayama, Tatsuya.

1992-01-01

The present invention rapidly detects leakage of primary coolants due to rupture of heat transfer pipes of a steam generator in a PWR type reactor to diagnose the operation state of the reactor. That is, a radiation detector is disposed to a secondary main steam pipeline for supplying steams generated from the steam generator to a turbine. The radiation detector detects a dose rate or a counting rate continuously. The measured data are transferred to an calculation and processing system and compared with the standard of normal values to diagnose the presence of leaks. Alternatively, radiation detectors are disposed at the upstream and the downstream of the secondary system main steam pipeline respectively. The signals from each of the radiation detectors are processed by the calculation and processing system as the change with lapse of time. As a result, the scale of the ruptured portion of the heat transfer pipe in the steam generator is diagnosed based on the value of radioactivity concentration in the main steams. (I.S.)

Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

International Nuclear Information System (INIS)

Woo, H.H.; Lu, S.C.

1981-01-01

Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design

Analysis of ex-vessel steam explosion with MC3D

International Nuclear Information System (INIS)

Leskovar, M.; Mavko, B.

2007-01-01

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which was developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central, left and right side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. The performed analysis shows that for some ex-vessel steam explosion scenarios significantly higher pressure loads are predicted than obtained in the OECD programme SERENA Phase 1. (author)

Evaluation on mechanical and corrosion properties of steam generator tubing materials

International Nuclear Information System (INIS)

Kim, In Sup; Lee, Byong Whi; Lee, Sang Kyu; Lee, Young Ho; Kim, Jun Whan; Lee, Ju Seok; Kwon, Hyuk Sang; Kim, Su Jung

1998-06-01

Steam generator is one of the major components of nuclear reactor pressure boundary. It's main function os transferring heat which generated in the reactor to turbine generator through steam generator tube. In these days, steam generator tubing materials of operating plant are used Inconel 600 alloys. But according to the operation time, there are many degradation phenomena which included mechanical damage due to flow induced vibration and corrosion damage due to PWSCC, IGA/SCC and pitting etc. Recently Inconel 690 alloys are selected as new and replacement steam generator tubes for domestic nuclear power plant. But there are few study about mechanical and corrosion properties of Inconel 600 and 690. The objectives of this study is to evaluate and compare mechanical and corrosion propertied of steam generator tube materials

Evaluation on mechanical and corrosion properties of steam generator tubing materials

Energy Technology Data Exchange (ETDEWEB)

Kim, In Sup; Lee, Byong Whi; Lee, Sang Kyu; Lee, Young Ho; Kim, Jun Whan; Lee, Ju Seok; Kwon, Hyuk Sang; Kim, Su Jung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-06-15

Steam generator is one of the major components of nuclear reactor pressure boundary. It's main function os transferring heat which generated in the reactor to turbine generator through steam generator tube. In these days, steam generator tubing materials of operating plant are used Inconel 600 alloys. But according to the operation time, there are many degradation phenomena which included mechanical damage due to flow induced vibration and corrosion damage due to PWSCC, IGA/SCC and pitting etc. Recently Inconel 690 alloys are selected as new and replacement steam generator tubes for domestic nuclear power plant. But there are few study about mechanical and corrosion properties of Inconel 600 and 690. The objectives of this study is to evaluate and compare mechanical and corrosion propertied of steam generator tube materials.

Method of judging leak sources in a reactor container

International Nuclear Information System (INIS)

Maeda, Katsuji.

1984-01-01

Purpose: To enable exact judgement for leak sources upon leak accident in a reactor container of BWR type power plants as to whether the sources are present in the steam system or coolant system. Method: If leak is resulted from the main steam system, the hydrogen density in the reactor container is about 170 times as high as the same amount of leak from the reactor water. Accordingly, it can be judged whether the leak source is present in the steam system or reactor water system based on the change in the indication of hydrogen densitometer within the reactor container, and the indication from the drain amount from the sump in the container or the indication of a drain flow meter in the container dehumidifier. Further, I-131, Na-24 and the like as the radioactive nucleides in sump water of the container are measured to determine the density ratio R = (I-131)/(Na-24), and it is judged that the leak is resulted in nuclear water if the density ratio R is equal to that of reactor water and that the leak is resulted from the main steam or like other steam system if the density ratio R is higher than by about 100 times than that of reactor water. (Horiuchi, T.)

Steam cleaning device

International Nuclear Information System (INIS)

Karaki, Mikio; Muraoka, Shoichi.

1985-01-01

Purpose: To clean complicated and long objects to be cleaned having a structure like that of nuclear reactor fuel assembly. Constitution: Steams are blown from the bottom of a fuel assembly and soon condensated initially at the bottom of a vertical water tank due to water filled therein. Then, since water in the tank is warmed nearly to the saturation temperature, purified water is supplied from a injection device below to the injection device above the water tank on every device. In this way, since purified water is sprayed successively from below to above and steams are condensated in each of the places, the entire fuel assembly elongated in the vertical direction can be cleaned completely. Water in the reservoir goes upward like the steam flow and is drained together with the eliminated contaminations through an overflow pipe. After the cleaning has been completed, a main steam valve is closed and the drain valve is opened to drain water. (Kawakami, Y.)

Study of steam, helium and supercritical CO2 turbine power generations in prototype fusion power reactor

International Nuclear Information System (INIS)

Ishiyama, Shintaro; Muto, Yasushi; Kato, Yasuyoshi; Nishio, Satoshi; Hayashi, Takumi; Nomoto, Yasunobu

2008-01-01

Power generation systems such as steam turbine cycle, helium turbine cycle and supercritical CO 2 (S-CO 2 ) turbine cycle are examined for the prototype nuclear fusion reactor. Their achievable cycle thermal efficiencies are revealed to be 40%, 34% and 42% levels for the heat source outlet coolant temperature of 480degC, respectively, if no other restriction is imposed. In the current technology, however, low temperature divertor heat source is included. In this actual case, the steam turbine system and the S-CO 2 turbine system were compared in the light of cycle efficiency and plant cost. The values of cycle efficiency were 37.7% and 36.4% for the steam cycle and S-CO 2 cycle, respectively. The construction cost was estimated by means of component volume. The volume became 16,590 m 3 and 7240 m 3 for the steam turbine system and S-CO 2 turbine system, respectively. In addition, separation of permeated tritium from the coolant is much easier in S-CO 2 than in H 2 O. Therefore, the S-CO 2 turbine system is recommended to the fusion reactor system than the steam turbine system. (author)

Large steam turbines for nuclear power stations. Output growth prospects

International Nuclear Information System (INIS)

Riollet, G.; Widmer, M.; Tessier, J.

1975-01-01

The rapid growth of the output of nuclear reactors, even if temporary settlement occurs, leads the manufacturer to evaluate, at a given time, technological limitations encountered. The problems dealing with the main components of turbines: steam path, rotors and stators steam valves, controle devices, shafts and bearings, are reviewed [fr

TRIGA reactor main systems

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2007-01-01

This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

Nuclear reactor

International Nuclear Information System (INIS)

Miyashita, Akio.

1981-01-01

Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

Detection of steam generator tube leaks in pressurized water reactors

International Nuclear Information System (INIS)

Roach, W.H.

1984-11-01

This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It identifies physical parameters, establishes instrumentation performance goals, and specifies sensor types and locations. It presents a simple algorithm that yields the leak rate as a function of known or measurable quantities. Leak rates of less than one-tenth gram per second should be detectable with existing instrumentation

Pressurizer and steam-generator behavior under PWR transient conditions

International Nuclear Information System (INIS)

Wahba, A.B.; Berta, V.T.; Pointner, W.

1983-01-01

Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

Steam reforming of heptane in a fluidized bed membrane reactor

Science.gov (United States)

Rakib, Mohammad A.; Grace, John R.; Lim, C. Jim; Elnashaie, Said S. E. H.

n-Heptane served as a model compound to study steam reforming of naphtha as an alternative feedstock to natural gas for production of pure hydrogen in a fluidized bed membrane reactor. Selective removal of hydrogen using Pd 77Ag 23 membrane panels shifted the equilibrium-limited reactions to greater conversion of the hydrocarbons and lower yields of methane, an intermediate product. Experiments were conducted with no membranes, with one membrane panel, and with six panels along the height of the reactor to understand the performance improvement due to hydrogen removal in a reactor where catalyst particles were fluidized. Results indicate that a fluidized bed membrane reactor (FBMR) can provide a compact reformer for pure hydrogen production from a liquid hydrocarbon feedstock at moderate temperatures (475-550 °C). Under the experimental conditions investigated, the maximum achieved yield of pure hydrogen was 14.7 moles of pure hydrogen per mole of heptane fed.

Effects of the Pressurized Water Reactor Main Steam Line Break Location on the Blowdown Loading

International Nuclear Information System (INIS)

Jo, Jong Chull; Kang, Soon Ho; Chan, Won Joon

2016-01-01

The thermal hydraulic analysis has been performed generally using a simple lumped model or one dimensional numerical model. However, those models have limitations in predicting the transient variations of the steam velocity, pressure and hydrodynamic load at a local point and the most conservative condition. Furthermore, it cannot be confirmed if the blowdown loads predicted by either of the models are conservative to evaluate every part of the SG internal structures. In this study, the transient hydraulic response of the SG secondary side to the MSLB case where the pipe break is assumed to occur at the SG outlet nozzle end, another weld point on the MSL, was numerically simulated using a CFD code. The present CFD calculation results was compared with those in ref. to investigate the effect of break location (friction loss) on the blowdown load in the SG secondary side. The result shows that the friction loss along the steam line span between the SG nozzle end and the MSIV would cause reduction in steam velocity disturbance or dynamic pressure. It implies that the consequence of the MSLB at the SG nozzle end would be much severer that those of other MSLB cases where the break locations are far from the SG. Therefore, to assure a conservative safety evaluation of the SG structural integrity, the blowdown loading on the SG internal structures including tubes during a MSLB accident in terms of the transient steam velocity, dynamic pressure and decompression wave fluctuations should be assessed for the MSLB case where the break is assumed to occur at the SG nozzle end.

Effects of the Pressurized Water Reactor Main Steam Line Break Location on the Blowdown Loading

Energy Technology Data Exchange (ETDEWEB)

Jo, Jong Chull; Kang, Soon Ho; Chan, Won Joon [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The thermal hydraulic analysis has been performed generally using a simple lumped model or one dimensional numerical model. However, those models have limitations in predicting the transient variations of the steam velocity, pressure and hydrodynamic load at a local point and the most conservative condition. Furthermore, it cannot be confirmed if the blowdown loads predicted by either of the models are conservative to evaluate every part of the SG internal structures. In this study, the transient hydraulic response of the SG secondary side to the MSLB case where the pipe break is assumed to occur at the SG outlet nozzle end, another weld point on the MSL, was numerically simulated using a CFD code. The present CFD calculation results was compared with those in ref. to investigate the effect of break location (friction loss) on the blowdown load in the SG secondary side. The result shows that the friction loss along the steam line span between the SG nozzle end and the MSIV would cause reduction in steam velocity disturbance or dynamic pressure. It implies that the consequence of the MSLB at the SG nozzle end would be much severer that those of other MSLB cases where the break locations are far from the SG. Therefore, to assure a conservative safety evaluation of the SG structural integrity, the blowdown loading on the SG internal structures including tubes during a MSLB accident in terms of the transient steam velocity, dynamic pressure and decompression wave fluctuations should be assessed for the MSLB case where the break is assumed to occur at the SG nozzle end.

High-definition radiography of tube-to-tubesheet welds of steam generator of prototype fast breeder reactor

International Nuclear Information System (INIS)

Venkatraman, B.; Sethi, V.K.; Jayakumar, T.; Raj, B.

1995-01-01

In the steam generator of the Prototype Fast Breeder Reactor (PFBR), steam is generated by the transfer of heat from secondary sodium to water. Due to the inherent dangers of sodium-water reaction, the integrity of weld joints separating sodium and water/steam is of paramount importance. This is particularly true and very important for the tube-to-tubesheet joints. This paper discusses the use of projective magnification technique by microfocal radiography for the quality evaluation and optimisation of the welding parameters of such small tube-to-tubesheet welds of the steam generator of PFBR. (author)

Steam--water mixing in nuclear reactor safety loss-of-coolant experiments

International Nuclear Information System (INIS)

Naff, S.A.; Schwarz, W.F.

1978-01-01

Computer models used to predict the response of reactors to hypothesized accidents necessarily incorporate approximating assumptions. To verify the models by comparing predicted and measured responses in test facilities, these assumptions must be confirmed to be realistic. Recent experiments in facilities capable of repeatedly duplicating the transient behavior of a pressurized water reactor undergoing a pipe rupture show that the assumption of complete water-steam mixing during the transient results in the predicted decompression being faster than that observed. Water reactor safety studies currently in progress include programs aimed at the verification of computer models or ''codes'' used to predict reactor system responses to various hypothesized accidents. The approach is to compare code predictions of transients with the actual test transients in experimental facilities. The purpose of this paper is to explain an important instance in which predictions and data are not in complete agreement and to indicate the significance to water reactor safety studies

Design and construction of past and present steam generators for the UK fast reactors

International Nuclear Information System (INIS)

Hayden, O.

1975-01-01

Double barrier sodium/water steam raising units were incorporated into the early DFR which has been operating since 1958, but to be economically viable the PFR units had to adopt a single wall concept. It should be remembered that the design style of PFR was decided upon in the early 1960's, when a very cautious approach had to be made. It was vital to ensure that a steam raising unit had the maximum availability and so a forced circulation system was chosen, making the steam generators consistent with all other power station boilers installed in the U.K. at that time. It is only recently that once-through steam cycles have been accepted in this country by the CEGB and these are on the A.G.R. stations currently being built. The 250 MWe Prototype Fast Reactor incorporates the world's largest sodium heated boilers and having gone critical some 6 months ago is currently undergoing various commissioning trials prior to its run-up to full power. The paper gives a brief description of these, with comments on particular features of Design, Development, nd countered to date are discussed, together with the way in which these have been overcome. Extensive research and development work has been carried out in support of Prototype Reactor and some of this continues well into the manufacture and commissioning programme. Critical areas such as tube-to-tubesheet welding, tube-grid fretting, burst disc and sodium/ water reaction work involves costly and time-consuming development. Sodium heated steam generators pose a greater number of problems to the designer than either fossil-fired or other types of nuclear steam raising plant. As in any boiler, economics demand that the heat exchanger surface is as compact as possible whilst retaining good 'low distribution inside and outside the tubes. Besides achieving a good thermal and mechanical design, the designer is faced with the possibility of a leak occurring and has to cater for the sodium/water reaction which may follow. Valuable

Design and construction of past and present steam generators for the UK fast reactors

Energy Technology Data Exchange (ETDEWEB)

Hayden, O

1975-07-01

Double barrier sodium/water steam raising units were incorporated into the early DFR which has been operating since 1958, but to be economically viable the PFR units had to adopt a single wall concept. It should be remembered that the design style of PFR was decided upon in the early 1960's, when a very cautious approach had to be made. It was vital to ensure that a steam raising unit had the maximum availability and so a forced circulation system was chosen, making the steam generators consistent with all other power station boilers installed in the U.K. at that time. It is only recently that once-through steam cycles have been accepted in this country by the CEGB and these are on the A.G.R. stations currently being built. The 250 MWe Prototype Fast Reactor incorporates the world's largest sodium heated boilers and having gone critical some 6 months ago is currently undergoing various commissioning trials prior to its run-up to full power. The paper gives a brief description of these, with comments on particular features of Design, Development, nd countered to date are discussed, together with the way in which these have been overcome. Extensive research and development work has been carried out in support of Prototype Reactor and some of this continues well into the manufacture and commissioning programme. Critical areas such as tube-to-tubesheet welding, tube-grid fretting, burst disc and sodium/ water reaction work involves costly and time-consuming development. Sodium heated steam generators pose a greater number of problems to the designer than either fossil-fired or other types of nuclear steam raising plant. As in any boiler, economics demand that the heat exchanger surface is as compact as possible whilst retaining good 'low distribution inside and outside the tubes. Besides achieving a good thermal and mechanical design, the designer is faced with the possibility of a leak occurring and has to cater for the sodium/water reaction which may follow. Valuable

Advanced boiling water reactor

International Nuclear Information System (INIS)

Nishimura, N.; Nakai, H.; Ross, M.A.

1999-01-01

In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

Thermal hydraulic aspects of steam drum level control philosophy for the natural circulation based heavy water reactor

International Nuclear Information System (INIS)

Gupta, S.K.; Gaikwad, A.J.; Kumar, Rajesh

2004-01-01

From safety considerations advanced nuclear reactors rely more and more on passive systems such as natural circulation for primary heat removal. A natural circulation based water reactor is relatively larger in size so as to reduce flow losses and channel type for proper flow distribution. From the size of steam drum considerations it has to be multi loop but has a common inlet header. Normally the turbine follows the reactor. This paper addresses the thermal hydraulic aspects of the steam drum pressure and level control philosophy for a four drum, natural circulation based, channel type boiling water advanced reactor. Three philosophies may be followed for drum control viz. individual drum control, one control drum approach and an average of all the four drums. For drum pressure control, the steam flow to the turbine is be regulated. A single point pressure control is better than individual drum pressure control. This is discussed in the paper. But the control point has to be at a place down steam the point where all steam line from individual drum meet. This may lead to different pressure in all the four drums depending on the power produced in the respective loops. The difference in pressure cannot be removed even if the four drums are directly connected through pipes. Also the pressure control scheme with/without interconnection is discussed. For level, the control of individual drum may not be normally possible because of common inlet header. As the frictional pressure drops in the large diameter downcomers are small as compared to elevation pressure drops, the level in all the steam drum tend to equalize. Consequently a single representative drum level may be chosen as a control variable for controlling level in all the four drums. But in case, where all the four loops are producing different powers and single point pressure control is effective, the scheme may not work satisfactorily. the level in a drum may depend on the power produced in the loop

Characteristics of fluctuating pressure generated in BWR main steam lines

International Nuclear Information System (INIS)

Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

2009-01-01

The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

A system for regulating the pressure of resuperheated steam in high temperature gas-cooled reactor power stations

International Nuclear Information System (INIS)

Braytenbah, A.S.; Jaegines, K.O.

1975-01-01

The invention relates to a system for regulating steam-pressure in the re-superheating portion of a steam-boiler receiving heat from a gas-cooled high temperature nuclear reactor, provided with gas distributing pumps driven by steam-turbines. The system comprises means for generating a pressure signal of desired magnitude for the re-superheating portion, and means for providing a real pressure in the re-superheating portion, means (including a by-passing device) for generating steam-flow rate signal of desired magnitude, a turbine by-pass device comprising a by-pass tapping means for regulating the steam-flow-rate in said turbine according to the desired steam-flow rate signal and means for controlling said by-pass tapping means according to said desired steam-flow-rate signal [fr

Numerical simulations of pressure fluctuations at branch piping in BWR main steam line

International Nuclear Information System (INIS)

Morita, Ryo; Inada, Fumio; Yoshikawa, Kazuhiro; Takahashi, Shiro

2009-01-01

The power uprating of a nuclear power plant may increase/accelerate degradation phenomena such as flow-induced vibration and wall thinking. A steam dryer was damaged by a high cycle fatigue due to an acoustic-induced vibration at the branch piping of safety relief valves (SRVs) in main steam lines. In this study, we conducted the numerical simulations of steam/air flow around a simplified branch piping to clarify the basic characteristics of resonance. LES simulations were conducted in ordinary pressure/temperature air and steam under BWR plant conditions. In both cases, the excitation of the pressure fluctuations at the branch was observed under some inlet velocity conditions. These fluctuations and inlet conditions were normalized and the obtained results were compared. The normalized results showed that the range and maximum amplitude of pressure fluctuations were almost the same in low-pressure/temperature air and high-pressure/temperature steam. We found that ordinary pressure/temperature air experiments and simulations can possibly clarify the characteristics of the resonance in high-pressure/temperature steam. (author)

Steam condenser optimization using Real-parameter Genetic Algorithm for Prototype Fast Breeder Reactor

Energy Technology Data Exchange (ETDEWEB)

Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Kumar, L. Satish, E-mail: satish@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Jehadeesan, R., E-mail: jeha@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Rajeswari, S., E-mail: raj@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Satya Murty, S.A.V., E-mail: satya@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India); Balasubramaniyan, V.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, Tamil Nadu (India)

2011-10-15

Highlights: > We model design optimization of a vital reactor component using Genetic Algorithm. > Real-parameter Genetic Algorithm is used for steam condenser optimization study. > Comparison analysis done with various Genetic Algorithm related mechanisms. > The results obtained are validated with the reference study results. - Abstract: This work explores the use of Real-parameter Genetic Algorithm and analyses its performance in the steam condenser (or Circulating Water System) optimization study of a 500 MW fast breeder nuclear reactor. Choice of optimum design parameters for condenser for a power plant from among a large number of technically viable combination is a complex task. This is primarily due to the conflicting nature of the economic implications of the different system parameters for maximizing the capitalized profit. In order to find the optimum design parameters a Real-parameter Genetic Algorithm model is developed and applied. The results obtained are validated with the reference study results.

Steam generator performance improvements for integral small modular reactors

Directory of Open Access Journals (Sweden)

Muhammad Ilyas

2017-12-01

Results and Conclusions: The results are compared with helical-coiled SGs being used in IRIS (International Reactor Innovative and Secure. The results show that the tube length reduces up to 11.56% keeping thermal and hydraulic conditions fixed. In the case of fixed size, the steam outlet temperature increases from 590.1 K to 597.0 K and the capability of power transfer from primary to secondary also increases. However, these advantages are associated with some extra pressure drop, which has to be compensated.

The study of steam explosions in nuclear systems. Advanced Reactor Severe Accident Program

International Nuclear Information System (INIS)

Theofanous, T.G.; Yuen, W.W.; Angelini, S.; Chen, X.

1995-01-01

This report presents an overview of the steam explosion issue in nuclear reactor safety and our approach to assessing it. Key physics, models, and computational tools are described, and illustrative results are presented for ex-vessel steam explosions in an open pool geometry. An extensive set of appendices facilitate access to previously reported work that is an integral part of this effort. These appendices include key developments in our approach, key advances in our understanding from physical and numerical experiments, and details of the most advanced computational results presented in this report. Of major significance are the following features: A consistent two-dimensional treatment for both premixing and propagation which in practical settings are ostensibly at least two-dimensional phenomena; experimental demonstration of voiding and microinteractions which represent key behaviors in premixing and propagation respectively; demonstration of the explosion venting phenomena in open pool geometries which, therefore, can be counted on as a very important mitigative feature; and introduction of the idea of penetration cutoff as a key mechanism prohibiting large-scale premixing in usual ex-vessel situations involving high pour velocities and subcooled pools. This report is intended as an overview and is to be followed by code manuals for PM-ALPHA and ESPROSE.m, respective verification reports, and application documents for reactor-specific applications. The applications will employ the Risk Oriented Accident Analysis Methodology (ROAAM) to address the safety importance of potential steam explosions phenomena in evaluated severe accidents for passive Advanced Light Water Reactors (ALWRs)

Three-dimensional neutron kinetics-thermal-hydraulics VVER 1000 main steam line break analysis by RELAP5-3D code

International Nuclear Information System (INIS)

Frisani, A.; Parisi, C.; D'Auria, F.

2007-01-01

After the development and the assessment of Three-Dimensional (3D) Neutron Kinetics (NK) - 1D Thermal-Hydraulics (TH) coupled codes analyses methods, deterministic nuclear safety technology is nowadays producing noticeable efforts for the validation of 3D NK - 3D TH coupled codes analyses methods too. Thus, the purpose of this work was to address the capability of the RELAP5-3D 3D NK-3D TH code to reproduce VVER 1000 Nuclear Power Plant (NPP) core dynamic in simulating the mixing effects that could happen in the vessel downcomer and lower plenum during some scenarios. The work was developed in three steps. The first step dealt with the 3D TH modeling of the Kozloduy-6 VVER 1000 reactor pressure vessel. Then this model was validated following a Steam Generator Isolation transient. The second step has been the development of a 3D NK nodalization for the reactor core region. Then the 3D NK model was directly coupled with the previously developed 3D TH model. The third step was the calculation of a Main Steam Line Break (MSLB) transient. The 3D NK global nuclear parameters were then compared with the 0-D results showing a good agreement; nevertheless only the 3D NK- 3D TH model allowed the calculation of each single assembly power trend for this strong NK-TH asymmetric transient. (author)

A nodalization study of steam separator in real time simulation

International Nuclear Information System (INIS)

Horugshyang, Lein; Luh, R.T.J.; Zen-Yow, Wang

1999-01-01

The motive of this paper is to investigate the influence of steam separator nodalization on reactor thermohydraulics in terms of stability and level response. Three different nodalizations of steam separator are studied by using THEATRE and REMARK Code in a BWR simulator. The first nodalization is the traditional one with two nodes for steam separator. In this nodalization, the steam separation is modeled in the outer node, i.e., upper downcomer. Separated steam enters the Steen dome node and the liquid goes to the feedwater node. The second nodalization is similar to the first one with the steam separation modeled in the inner node. There is one additional junction connecting steam dome node and the inner node. The liquid fallback junction connects the inner node and feedwater node. The third nodalization is a combination of the former two with an integrated node for steam separator. Boundary conditions in this study are provided by a simplified feedwater and main steam driver. For comparison purpose, three tests including full power steady state initialisation, recirculation pumps runback and reactor scram are conducted. Major parameters such as reactor pressure, reactor level, void fractions, neutronic power and junction flows are recorded for analysis. Test results clearly show that the first nodalization is stable for steady state initialisation. However it has too responsive level performance in core flow reduction transients. The second nodalization is the closest representation of real plant structure, but not the performance. Test results show that an instability occurs in the separator region for both steady state initialisation and transients. This instability is caused by an unbalanced momentum in the dual loop configuration. The magnitude of the oscillation reduces as the power decreases. No superiority to the other nodalizations is shown in the test results. The third nodalization shows both stability and responsiveness in the tests. (author)

Control device for start-up of reactor depressurization system

International Nuclear Information System (INIS)

Suzuki, Hiroshi; Saito, Minoru; Oda, Shingo; Miura, Satoshi; Hashimoto, Koji; Tate, Hitoshi; Fujii, Kazunobu

1998-01-01

The present invention concerns are emergency reactor core cooling system (ECCS) of a BWR type reactor and provides a control device for start-up of an automatic depressurization system. Namely, the device has an object of preventing erroneous opening of a main steam escape safety value when testing a start-up signal circuit of an automatic depressurization system for testing the automatic depressurization system. A start-up signal circuit receives both signals of a reactor container pressure high signal and a reactor pressure vessel water level low signal and outputs an automatic start-up signal for compulsorily opening a main steam escape safety valve automatically. A test switch having a self-holding circuit is disposed to a central control chamber. A test signal circuit is disposed for preventing transfer of an erroneous start-up signal to the main steam escape safety valve due to a simulation signal during output test signals by the test switch. (I.S.)

Experience in adjusting of the level regulation system of steam generators of the Rovno NPP

International Nuclear Information System (INIS)

Patselyuk, S.N.; Sokolov, A.G.; Kazakov, V.I.; Dorosh, Yu.A.

1984-01-01

A system of feed water level control in steam generators at the Rovno NPP with WWER-440 reactors which comprises start-up as well as main regulators is described. The start-up regulator (single-pulsed with a signal by the level) keeps the level in the steam generator at loadings up to 30% of the nominal reactor power Nsub(nom.) The main regulator is connected in the three-pulsed circuit and it receives signals by steam and water flow rate and by the level in the steam generator. The main regulator has been started only at loadings above 40% Nsub(nom.). After reconstruction it was used in the 15-100% Nsub(nom.) range. Characteristics of the level control system in the steam generator at perturbations intoduced by the main circulating pump (MCP) and turbine disconnection as well as change in feed water flow rate have been studied. The studies have revealed that the system ensures necessary quality of control in stationary modes. The system operates stably at perturbations of feed water flow rate up to 50% Nsub(nom.). Perturbations by MCP connections and disconnections is most difficult for control system

Thermoelectric generation coupling methanol steam reforming characteristic in microreactor

International Nuclear Information System (INIS)

Wang, Feng; Cao, Yiding; Wang, Guoqiang

2015-01-01

Thermoelectric (TE) generator converts heat to electric energy by thermoelectric material. However, heat removal on the cold side of the generator represents a serious challenge. To address this problem and for improved energy conversion, a thermoelectric generation process coupled with methanol steam reforming (SR) for hydrogen production is designed and analyzed in this paper. Experimental study on the cold spot character in a micro-reactor with monolayer catalyst bed is first carried out to understand the endothermic nature of the reforming as the thermoelectric cold side. A novel methanol steam reforming micro-reactor heated by waste heat or methanol catalytic combustion for hydrogen production coupled with a thermoelectric generation module is then simulated. Results show that the cold spot effect exists in the catalyst bed under all conditions, and the associated temperature difference first increases and then decreases with the inlet temperature. In the micro-reactor, the temperature difference between the reforming and heating channel outlets decreases rapidly with an increase in thermoelectric material's conductivity coefficient. However, methanol conversion at the reforming outlet is mainly affected by the reactor inlet temperature; while at the combustion outlet, it is mainly affected by the reactor inlet velocity. Due to the strong endothermic effect of the methanol steam reforming, heat supply of both kinds cannot balance the heat needed at reactor local areas, resulting in the cold spot at the reactor inlet. When the temperature difference between the thermoelectric module's hot and cold sides is 22 K, the generator can achieve an output voltage of 55 mV. The corresponding molar fraction of hydrogen can reach about 62.6%, which corresponds to methanol conversion rate of 72.6%. - Highlights: • Cold spot character of methanol steam reforming was studied through experiment. • Thermoelectric generation Coupling MSR process has been

Optimization of a Pd-based membrane reactor for hydrogen production from methane steam reforming

Energy Technology Data Exchange (ETDEWEB)

Assis, A.J.; Hori, C.E.; Silva, L.C.; Murata, V.V. [Universidade Federal de Uberlandia (UFU), MG (Brazil). School of Chemical Engineering]. E-mail: adilsonjassis@gmail.com

2008-07-01

In this work, it is proposed a phenomenological model in steady state to describe the performance of a membrane reactor for hydrogen production through methane steam reform as well as it is performed an optimization of operating conditions. The model is composed by a set of ordinary differential equations from mass, energy and momentum balances and constitutive relations. They were used two different intrinsic kinetic expressions from literature. The results predicted by the model were validated using experimental data. They were investigated the effect of five important process parameters, inlet reactor pressure (PR0), methane feed flow rate (FCH40), sweep gas flow rate (FI), external reactor temperature (TW) and steam to methane feed flow ratio (M), both on methane conversion (XCH{sub 4} ) and hydrogen recovery (YH{sub 2}). The best operating conditions were obtained through simple parametric optimization and by a method based on gradient, which uses the computer code DIRCOL in FORTRAN. It is shown that high methane conversion (96%) as well as hydrogen recovery (91%) can be obtained, using the optimized conditions. (author)

The supply of steam from Candu reactors for heavy water production

International Nuclear Information System (INIS)

Robertson, R.F.S.

1975-09-01

By 1980, Canada's energy needs for D 2 O production will be 420 MW of electrical energy and 3600 MW of thermal energy (as steam). The nature of the process demands that this energy supply be exceptionally stable. Today, production plants are located at or close to nuclear electricity generating sites where advantage can be taken of the low cost of both the electricity and steam produced by nuclear reactors. Reliability of energy supply is achieved by dividing the load between the multiple units which comprise the sites. The present and proposed means of energy supply to the production sites at the Bruce Heavy Water Plant in Ontario and the La Prade Heavy Water Plant in Quebec are described. (author)

A charge regulating system for turbo-generator gas-cooled high-temperature reactor power stations

International Nuclear Information System (INIS)

Braytenbah, A.S.; Jaegtnes, K.O.

1975-01-01

The invention relates to a regulating system for gas-cooled high-temperature reactors power stations (helium coolant), equipped with several steam-boilers, each of which deriving heat from a corresponding cooling-gas flow circulating in the reactor, so as to feed superheated steam into a main common steam-manifold and re-superheated steam into a re-superheated hot common manifold [fr

AREVA Modular Steam Cycle – High Temperature Gas-Cooled Reactor Development Progress

International Nuclear Information System (INIS)

Lommers, L.; Shahrokhi, F.; Southworth, F.; Mayer, J. III

2014-01-01

The AREVA Steam Cycle – High Temperature Gas-Cooled Reactor (SCHTGR) is a modular graphite-moderated gas-cooled reactor currently being developed to support a wide variety of applications including industrial process heat, high efficiency electricity generation, and cogeneration. It produces high temperature superheated steam which makes it a good match for many markets currently dependent on fossil fuels for process heat. Moreover, the intrinsic safety characteristics of the SC-HTGR make it uniquely qualified for collocation with large industrial process heat users which is necessary for serving these markets. The NGNP Industry Alliance has selected the AREVA SC-HTGR as the basis for future development work to support commercial HTGR deployment. This paper provides a concise description of the SC-HTGR concept, followed by a summary of recent development activities. Since this concept was introduced, ongoing design activities have focused primarily on confirming key system capabilities and the suitability for potential future markets. These evaluations continue to confirm the suitability of the SC-HTGR for a variety of potential applications that are currently dependent on fossil fuels. (author)

Co-current and counter-current configurations for ethanol steam reforming in a dense Pd-Ag membrane reactor

NARCIS (Netherlands)

Gallucci, F.; de Falco, M.; Tosti, S.; Marrelli, L; Basile, A.

2008-01-01

The ethanol steam-reforming reaction to produce pure hydrogen has been studied theoretically. A mathematical model has been formulated for a traditional system and a palladium membrane reactor packed with a Co-based catalyst and the simulation results related to the membrane reactor for both

TRAC analysis of steam-generator overfill transients for TMI-1

International Nuclear Information System (INIS)

Bassett, B.

1983-01-01

A reactor safety issue concerning the overfilling of once-through steam generators leading to combined primary/secondary blowdown has been raised recently. A series of six calculations, performed with the LWR best-estimate code, TRAC-PD2, on a Babcock and Wilcox Plant (TMI-1), was performed to investigate this safety issue. The base calculation assumed runaway main feedwater to one steam generator causing it to overfill and to break the main steam line. Four additional calculations build onto the base case with combinations of a pump-seal failure, a steam-generator tube rupture, and the pilot-operated relief valve not reseating. A sixth calculation involved only the rupture of a single steam-generator tube. The results of these analyses indicate that for the transients investigated, the emergency cooling system provided an adequate make-up coolant flow to mitigate the accidents

Steam generator tube performance: experience with water-cooled nuclear power reactors during 1978

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1980-02-01

The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1978. Tube failures occurred at 31 of the 86 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. A dramatic decrease in the number of tubes plugged was evident in 1978 compared to the previous year. This is attributed to diligent application of techniques developed from in-plant experience and research and development programs over the past several years. (auth)

Steam generator tube performance: experience with water-cooled nuclear power reactors during 1977

International Nuclear Information System (INIS)

Pathania, R.S.; Tatone, O.S.

1979-02-01

The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1977. Failures were reported in 34 of the 79 reactors surveyed. Causes of these failures and inspection and repair procedures designed to deal with them are presented. Although corrosion remained the leading cause of tube failures, specific mechanisms have been identified and methods of dealing with them developed. These methods are being applied and should lead to a reduction of corrosion failures in future. (author)

Steam condenser optimization using Real-parameter Genetic Algorithm for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Jayalal, M.L.; Kumar, L. Satish; Jehadeesan, R.; Rajeswari, S.; Satya Murty, S.A.V.; Balasubramaniyan, V.; Chetal, S.C.

2011-01-01

Highlights: → We model design optimization of a vital reactor component using Genetic Algorithm. → Real-parameter Genetic Algorithm is used for steam condenser optimization study. → Comparison analysis done with various Genetic Algorithm related mechanisms. → The results obtained are validated with the reference study results. - Abstract: This work explores the use of Real-parameter Genetic Algorithm and analyses its performance in the steam condenser (or Circulating Water System) optimization study of a 500 MW fast breeder nuclear reactor. Choice of optimum design parameters for condenser for a power plant from among a large number of technically viable combination is a complex task. This is primarily due to the conflicting nature of the economic implications of the different system parameters for maximizing the capitalized profit. In order to find the optimum design parameters a Real-parameter Genetic Algorithm model is developed and applied. The results obtained are validated with the reference study results.

Design features of Advanced Power Reactor (APR) 1400 steam generator

International Nuclear Information System (INIS)

Park, Tae-Jung; Park, Jun-Soo; Kim, Moo-Yong

2004-01-01

Advanced Power Reactor 1400 (APR 1400) which is to achieve the improvement of the safety and economical efficiency has been developed by Korea Hydro and Nuclear Power Co., Ltd. (KHNP) with the support from industries and research institutes. The steam generator for APR 1400 is an evolutionary type from System 80 + , which is the recirculating U-tube heat exchanger with integral economizer. Compared to the System 80 + steam generator, it is focused on the improved design features, operating and design conditions of APR 1400 steam generator. Especially, from the operation experience of Korean Standard Nuclear Power Plant (KSNP) steam generator, the lessons-learned measures are incorporated to prevent the tube wear caused by flow-induced vibration (FIV). The concepts for the preventive design features against FIV are categorized to two fields; flow distribution and dynamic response characteristics. From the standpoint of flow distribution characteristics, the egg-crate flow distribution plate (EFDP) is installed to prevent the local excessive flow loaded on the most susceptible tube to wear. The parametric study is performed to select the optimum design with the efficient mitigation of local excessive flow. ATHOS3 Mod-01 is used and partly modified to analyze the flow field of the APR 1400 steam generator. In addition, the upper tube bundle support is designed to eliminate the presence of tube with a low natural frequency. Based on the improved upper tube bundle support, the modal analysis is performed and compared with that of System 80 + . Using the results of flow distribution and modal analysis, the two mechanisms of flow-induced vibration are investigated; fluid-elastic instability (FEI) and random turbulence excitation (RTE). (authors)

Steam-generator tube failures: world experience in water-cooled nuclear power reactors during 1972

International Nuclear Information System (INIS)

Stevens-Guille, P.D.

1975-01-01

During 1972, approximately one in three operating reactors with steam generators incurred tube failures, predominantly near the tube sheet and in the bend region. Various forms of corrosion were the most frequent cause of failure. Eddy-current inspection was the preferred method for locating and investigating the cause of failure. Extensive use was made of both mechanical and explosive plugs for repair. As a class, steam generators with Monel 400 tubes had the lowest failure rates, and those with Inconel 600 tubes had the highest. (U.S.)

Active acoustic leak detection in steam generator units of fast reactors

International Nuclear Information System (INIS)

Oriol, L.; Journeau, Ch.

1996-01-01

Steam generators (SG) of Fast Reactors can be subject to water leakage into the sodium secondary circuit, causing an exothermic chemical reaction with potential serious damage to plant. Within the framework of the European Fast Reactor project, the CEA has developed an active acoustic detection technique which, when used in parallel with passive acoustic detection, will lead to effective leak detection results in terms of reliability and false alarm rates. Whilst the passive method is based on the increase in acoustic noise generated by the reaction, the active method takes advantage of the acoustic attenuation by the hydrogen bubbles produced. The method has been validated: in water, during laboratory testing at the Centre d'Etudes de Cadarache; in sodium, at the ASB loop at Bensberg (Germany) and at AEA Dounreay (Scotland). Full analysis of the tests carried out on the SG of the Prototype Fast Reactor in 1994 during end-of-life testing should lead to reactor validation on the method. (authors)

Economic evaluation of the steam-cycle high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

1983-07-01

The High Temperature Gas-Cooled Reactor is unique among current nuclear technologies in its ability to generate energy in temperature regimes previously limited to fossil fuels. As a result, it can offer commercial benefits in the production of electricity, and at the same time, expand the role of nuclear energy to the production of process heat. This report provides an evaluation of the HTGR-Steam Cycle (SC) system for the production of baseloaded electricity, as well as cogenerated electricity and process steam. In each case the HTGR-SC system has been evaluated against appropriate competing technologies. The computer code which was developed for this evaluation can be used to present the analyses on a cost of production or cash flow basis; thereby, presenting consistent results to a utility, interested in production costs, or an industrial steam user or third party investor, interested in returns on equity. Basically, there are two economic evaluation methodologies which can be used in the analysis of a project: (1) minimum revenue requirements, and (2) discounted cash flow

Steam generator tube performance. Experience with water-cooled nuclear power reactors during 1985

International Nuclear Information System (INIS)

Tatone, O.S.; Tapping, R.L.

1988-12-01

The performance of steam generator tubes at water-cooled reactors during 1985 has been reviewed. Seventy-three of 168 reactors in the survey experienced tube degradation sufficient for the tubes to be plugged. The number of tubes plugged was 6837 or 0.28% of those in service. The leading cause of tube failure was stress corrosion cracking from the primary side. Stress corrosion cracking or intergranular attack from the secondary side and pitting were also major causes of tube failure. Unlike most previous years, fretting was a substantial problem at some reactors. Overall, corrosion continued to account for more than 80% of the defects. 20 refs

Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

Energy Technology Data Exchange (ETDEWEB)

Kral, P. [Nuclear Research Inst. Rez (Switzerland)

1995-12-31

The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

Energy Technology Data Exchange (ETDEWEB)

Kral, P [Nuclear Research Inst. Rez (Switzerland)

1996-12-31

The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

International Nuclear Information System (INIS)

Kral, P.

1995-01-01

The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal 'MSH Rupture' leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS

Auxiliary water supply device for BWR type reactor

International Nuclear Information System (INIS)

Sasagawa, Hiroshi.

1994-01-01

In the device of the present invention, a cooling condensation means is disposed to a steam discharge channel of a turbine for driving pumps to directly return condensates to the reactor, so that the temperature of the suppression pool water is not elevated. Namely, the cooling condensation means for discharged steams is disposed to the discharge channel of the turbine. The condensate channel from the cooling condensation means is connected to a sucking side of the turbine driving pump. With such a constitution, when the reactor is isolated from a main steam system, reactor scram is conducted. Although the reactor water level is lowered by the reactor scram, the lowering of the reactor water level is prevented by supplementing cooling water by the turbine driving pump using steams generated in the reactor as a power source. The discharged steams after driving the turbine are cooled and condensated by the cooling condensation means by way of the discharge channel and returned to the reactor again by way of the condensate channel. With such procedures, since the temperature of suppression pool water is not elevated, there is no need to operate other cooling systems. In addition, auxiliary water can be supplied for a long period of time. (I.S.)

Corrosion aspects in steam generators of nuclear power plants

International Nuclear Information System (INIS)

Visoni, E.; Santos Pinto, M. dos

1988-01-01

Steam generators of pressurized water reactors (PWR), transfer heat from a primary coolant system to a secondary coolant system. Primary coolant water is heated in the core and passes through the steam generator that transfer heat to the secondary coolant water. However, the steam generator is dead for ionic impurities, corrosion products and fabrication/maintenence residues. These impurities concentrate between crevice and cracks. Many types of degradation mechanisms affect the tubes. The tubes are dented, craked, ovalized, wasted, etc. This paper describes the main corrosion problems in steam generators and includes the corrective actions to considered to reduce or eliminate these corrosion problems. (author) [pt

Design and optimization of a fixed - bed reactor for hydrogen production via bio-ethanol steam reforming

International Nuclear Information System (INIS)

Maria A Goula; Olga A Bereketidou; Costas G Economopoulos; Olga A Bereketidou; Costas G Economopoulos

2006-01-01

Global climate changes caused by CO 2 emissions are currently debated around the world. Renewable sources of energy are being sought as alternatives to replace fossil fuels. Hydrogen is theoretically the best fuel, environmentally friendly and its combustion reaction leads only to the production of water. Bio-ethanol has been proven to be effective in the production of hydrogen via steam reforming reaction. In this research the steam reforming reaction of bio-ethanol is studied at low temperatures over 15,3 % Ni/La 2 O 3 catalyst. The reaction and kinetic analysis takes place in a fixed - bed reactor in 130 - 250 C in atmospheric pressure. This study lays emphasis on the design and the optimization of the fixed - bed reactor, including the total volume of the reactor, the number and length of the tubes and the degree of ethanol conversion. Finally, it is represented an approach of the total cost of the reactor, according to the design characteristics and the materials that can be used for its construction. (authors)

The steam explosion potential for an unseated SRS reactor septifoil

International Nuclear Information System (INIS)

Allison, D.K.; Hyder, M.L.; Yau, W.W.F.; Smith, D.C.

1992-01-01

Control rods in the Savannah River Site's K Reactor are contained within housings composed of seven channels (''septifoils''). Each septifoil is suspended from the top of the reactor and is normally seated on an upflow pin that channels coolant to the septifoil. Forced flow to the septifoil would be eliminated in the unlikely event of a septifoil unseated upon installation, i.e., if the septifoil is not aligned with its upflow pin. If this event were not detected, control rod melting and the interaction of molten metal with water might occur. This paper describes a methodology used to address the issue of steam explosions that might arise by this mechanism. The probability of occurrence of a damaging steam explosion given an unseated septifoil was found to be extremely low. The primary reasons are: (1) the high probability that melting will not occur, (2) the possibility of material holdup by contact with the outer septifoil housing, (3) the relative shallowness of the pool 'Of water into which molten material might fall, (4) the probable absence of a trigger, and (5) the relatively large energy release required to damage a nearby fuel assembly. The methodology is based upon the specification of conditions prevailing within the septifoil at the time molten material is expected to contact water, and upon information derived from the available experimental data base, supplemented by recent prototypic experiments

Theoretical-experimental modelling of the momentum equation for PWR reactor steam generators

International Nuclear Information System (INIS)

Rodrigues, L.A.H.

1994-01-01

A mathematical model in steady-state conditions of the momentum equation at the secondary side of a vertical U-tube steam generator with recirculation is presented. The U-tube test section was the 150 bar - Circuito Termoidraulico Experimental - CTE-150. This facility is a Experimental Thermal-hydraulic Circuit and operates at the same conditions (pressure and temperature) of a typical PWR reactor. A comparison between the Homogeneous and Separate Flow models was done. those models were verified and compared with experimental data for several operational conditions. The results show that the model fits very well the experimental data and seems to be appropriate to study water recirculation of a steam generator secondary side. (author)

Nuclear reactor plants and control systems therefor

International Nuclear Information System (INIS)

de Boer, G.A.; de Hex, M.

1976-01-01

A nuclear reactor plant is described comprising at least two hydraulically separated but thermally interconnected heat conveying circuits, of which one is the reactor circuit filled with a non-water medium and the other one is the water-steam-circuit equipped with a steam generator, a feed water conduit controlled by a valve and a steam turbine, and a control system mainly influenced by the pressure drop caused in said feed water conduit and its control valve and having a value of at least 10 bars at full load

Thermal expansion measurement of turbine and main steam piping by using strain gages in power plants

International Nuclear Information System (INIS)

Na, Sang Soo; Chung, Jae Won; Bong, Suk Kun; Jun, Dong Ki; Kim, Yun Suk

2000-01-01

One of the domestic co-generation plants have undergone excessive vibration problems of turbine attributed to external force for years. The root cause of turbine vibration may be shaft alignment problem which sometimes is changed by thermal expansion and external force, even if turbine technicians perfectly performed it. To evaluate the alignment condition from plant start-up to full load, a strain measurement of turbine and main steam piping subjected to thermal loading is monitored by using strain gages. The strain gages are bonded on both bearing housing adjusting bolts and pipe stoppers which installed in the x-direction of left-side main steam piping near the turbine inlet in order to monitor closely the effect of turbine under thermal deformation of turbine casing and main steam piping during plant full load. Also in situ load of constant support hangers in main steam piping system is measured by strain gages and its results are used to rebalance the hanger rod load. Consequently, the experimental stress analysis by using strain gages turns out to be very useful tool to diagnose the trouble and failures of not only to stationary components but to rotating machinery in power plants

A single-stage high pressure steam injector for next generation reactors: test results and analysis

International Nuclear Information System (INIS)

Cattadori, G.; Galbiati, L.; Mazzocchi, L.; Vanini, P.

1995-01-01

Steam injectors can be used in advanced light water reactors (ALWRs) for high pressure makeup water supply; this solution seems to be very attractive because of the ''passive'' features of steam injectors, that would take advantage of the available energy from primary steam without the introduction of any rotating machinery. The reference application considered in this work is a high pressure safety injection system for a BWR; a water flow rate of about 60 kg/s to be delivered against primary pressures covering a quite wide range up to 9 MPa is required. Nevertheless, steam driven water injectors with similar characteristics could be used to satisfy the high pressure core coolant makeup requirements of next generation PWRs. With regard to BWR application, an instrumented steam injector prototype with a flow rate scaling factor of about 1:6 has been built and tested. The tested steam injector operates at a constant inlet water pressure (about 0.2 MPa) and inlet water temperature ranging from 15 to 37 o C, with steam pressure ranging from 2.5 to 8.7 MPa, always fulfilling the discharge pressure target (10% higher than steam pressure). To achieve these results an original double-overflow flow rate-control/startup system has been developed. (Author)

Steam generator replacement in Bruce A Unit 1 and Unit 2

International Nuclear Information System (INIS)

Hart, R.S.

2007-01-01

The Bruce A Generating Station consists of four 900 MW class CANDU units. The reactor and Primary Heat Transport System for each Unit are housed within a reinforced concrete reactor vault. A large duct running below the reactor vaults accommodates the shared fuel handling system, and connects the four reactor vaults to the vacuum building. The reactor vaults, fuelling system duct and the vacuum building constitute the station vacuum containment system. Bruce A Unit 2 was shut down in 1995 and Bruce A Units 1, 3 and 4 were shutdown in 1997. Bruce A Units 3 and 4 were returned to service in late 2003 and are currently operating. Units 1 and 2 remain out of service. Bruce Power is currently undertaking a major rehabilitation of Bruce A Unit 1 and Unit 2 that will extend the in-service tile of these units by at least 25 years. Replacement of the Steam Generators (eight in each unit) is required; this work was awarded to SNC-Lavalin Nuclear (SLN). The existing steam drums (which house the steam separation and drying equipment) will be retained. Unit 2 is scheduled to be synchronized with the grid in 2009, followed by Unit 1 in 2009. Each Bruce A unit has two steam generating assemblies, one located above and to each end of the reactor. Each steam generating assembly consists of a horizontal cylindrical steam drum and four vertical Steam Generators. The vertical Steam Generators connect to individual nozzles that are located on the underside of the Steam Drum (SD). The steam drums are located in concrete shielding structures (steam drum enclosures). The lower sections of the Steam Generators penetrate the top of the reactor vaults: the containment pressure boundary is established by bellows assemblies that connect between the reactor vault roof slab and the Steam Generators. Each Steam Generators is supported from he bottom by a trapeze that is suspended from the reactor vault top structure. The Steam Generator Replacement (SGR) methodology developed by SLN for Unit 1

Investigations to the potential of the high temperature reactor for steam power processes with highest steam conditions and comparison with according conventional power plants

International Nuclear Information System (INIS)

Mondry, M.

1988-04-01

Already in the fifties conventional power plants with high parameters of the live steam were built to improve the total efficiency. The power plant with the highest steam conditions in the Federal Republic of Germany has 300 bar pressure and 600deg C temperature. Because of high material costs and other problems power plants with such high conditions were not continued to be built. Standard conditions of today's power plants are in the order of 180-250 bar pressure and 535deg C temperature. As the high temperature reactor is partly built up in another way than a conventional power plant, the results regarding the high steam parameters are not transferable. Possibilities for the technical realization of determined HTR-specific components are introduced and discussed. Then different HTR-power plants with steam conditions up to 350 bar pressure and 650deg C temperature are projected. Economical considerations show that an HTR with higher steam parameters brings financial profits. Further efficiency increase, which is possible by the high steam conditions, is shortly presented. The work ends with a technical and economical comparison of corresponding conventional power plants. (orig./UA) [de

Determination of stresses caused by fluctuation of acoustic load in the steam dryers of a BWR; Determinacion de esfuerzos originados por fluctuacion de carga acustica en los secadores de vapor de un reactor BWR

Energy Technology Data Exchange (ETDEWEB)

Centeno P, J.; Quezada G, S.; Prieto G, A.; Vazquez R, A.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico); Nunez C, A., E-mail: javcuami26@hotmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2014-10-15

The extended power up-rate (EPU) in a nuclear power plant cause various problems in BWR components also in the steam system. This due to increased steam flow generated in the reactor and is conveyed to the turbine by the four main steam lines (MSL). One of the most serious problems is the generation of acoustic pressure loads in the metal structure of the steam dryer which eventually leads to fatigue failure and even the appearance of cracks, and in turn it causes loose parts that are entrained by the steam and transported in the MSL. This problem is due to the fluctuation of load acoustics caused by the union of the safety or relief valves (SRV) with the MSL, spreading through these to reach the reactor pressure vessel (RPV) where the effect of resonance of the acoustic wave is amplified and impacts directly in the supporting structure of the steam dryer, skirt and the panels where the mixture liquid-steam is dried, by centrifugation effect and runoff of liquid water. Efforts in the steam dryer operating conditions of EPU for two cases will be analyzed in this work, the first is before the installation of Acoustic Side Branch (ASB), and in the second case we consider the installation of said ASB in the standpipes of SRV. The analysis was performed with numerical experiments on a platform for computational fluid dynamics with virtual geometries previously designed based on the actual components of the reactor and steam system. The model to study is delimited by the top of the RPV, the steam dryer and a section of each of the four MSL with ten standpipes of SRV. With the obtained data and considering the mechanical-structural properties of the steam dryer material, we can evaluate the mechanical resistance to impacts by acoustic pressure load and its possible deformation or cracking. (Author)

Steam generator replacement in Bruce A Unit 1 and Unit 2

International Nuclear Information System (INIS)

Hart, R.S.

2006-01-01

The Bruce A Generating Station consists of four 900 MW class CANDU units. The reactor and Primary Heat Transport System for each Unit are housed within a reinforced concrete reactor vault. A large duct running below the reactor vaults accommodates the shared fuel handling system, and connects the four reactor vaults to the vacuum building. The reactor vaults, fuelling system duct and the vacuum building constitute the station vacuum containment system. Bruce A Unit 2 was shut down in 1995 and Bruce A Units 1, 3 and 4 were shutdown in 1997. Bruce A Units 3 and 4 were returned to service in late 2003 and are currently operating. Units 1 and 2 remain out of service. Bruce Power is currently undertaking a major rehabilitation of Bruce A Unit 1 and Units 2 that will extend the in-service life of these units by at least 25 years. Replacement of the Steam Generators (eight in each unit) is required; this work was awarded to SNC-Lavalin Nuclear (SLN). The existing steam drums (which house the steam separation and drying equipment) will be retained. Unit 2 is scheduled to be synchronized with the grid in 2009, followed by Unit 1 in 2009. Each Bruce A unit has two steam generating assemblies, one located above and to each end of the reactor. Each steam generating assembly consists of a horizontal cylindrical steam drum and four vertical Steam Generators. The vertical Steam Generators connect to individual nozzles that are located on the underside of the Steam Drum (SD). The steam drums are located in concrete shielding structures (steam drum enclosures). The lower sections of the Steam Generators penetrate the top of the reactor vaults: the containment pressure boundary is established by bellows assemblies that connect between the reactor vault roof slab and the Steam Generators. Each Steam Generators is supported from the bottom by a trapeze that is suspended from the reactor vault top structure. The Steam Generator Replacement (SGR) methodology developed by SLN for Unit 1

Steam Reformer With Fibrous Catalytic Combustor

Science.gov (United States)

Voecks, Gerald E.

1987-01-01

Proposed steam-reforming reactor derives heat from internal combustion on fibrous catalyst. Supplies of fuel and air to combustor controlled to meet demand for heat for steam-reforming reaction. Enables use of less expensive reactor-tube material by limiting temperature to value safe for material yet not so low as to reduce reactor efficiency.

A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

Energy Technology Data Exchange (ETDEWEB)

Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong [Korea Power Engineering Company, Inc, 150 Deokjin-dong, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2006-07-01

This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

International Nuclear Information System (INIS)

Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong

2006-01-01

This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

Advanced steam cycles for light water reactors. Final report

International Nuclear Information System (INIS)

Mitchell, R.C.

1975-07-01

An appraisal of the potential of adding superheat to improve the overall LWR plant cycle performance is presented. The study assesses the economic and technical problems associated with the addition of approximately 500 0 F of superheat to raise the steam temperature to 1000 0 F. The practicality of adding either nuclear or fossil superheat to LWR's is reviewed. The General Electric Company Boiling Water Reactor (BWR) model 238-732 (BWR/6) is chosen as the LWR starting point for this evaluation. The steam conditions of BWR/6 are representative of LWR's. The results of the fossil superheat portion of the evaluation are considered directly applicable to all LWR's. In spite of the potential of a nuclear superheater to provide a substantial boost to the LWR cycle efficiency, nuclear superheat offers little promise of development at this time. There are difficult technical problems to resolve in the areas of superheat fuel design and emergency core cooling. The absence of a developed high integrity, high temperature fuel for operation in the steam/water environment is fundamental to this conclusion. Fossil superheat offers the potential opportunity to utilize fossil fuel supplies more efficiently than in any other mode of central station power generation presently available. Fossil superheat topping cycles evaluated included atmospheric fluidized beds (AFB), pressurized fluidized beds, pressurized furnaces, conventional furnaces, and combined gas/steam turbine cycles. The use of an AFB is proposed as the preferred superheat furnace. Fossil superheat provides a cycle efficiency improvement for the LWR of two percentage points, reduces heat rejection by 15 percent per kWe generated, increases plant electrical output by 54 percent, and burns coal with an incremental net efficiency of approximately 40 percent. This compares with a net efficiency of 36--37 percent which might be achieved with an all-fluidized bed fossil superheat plant design

Ex-Vessel Steam Explosion Analysis of Central Melt Pour Scenario

International Nuclear Information System (INIS)

Ursic, M.; Leskovar, M.

2008-01-01

An ex-vessel steam explosion may develop during a severe reactor accident when the reactor vessel fails and the molten core interacts with the coolant in the reactor cavity. At this process part of the corium energy is intensively transferred to water in a very short time scale. The water vaporizes at high pressure and expands, doing work on its surrounding. Although the steam explosion has probably a low probability of occurrence, it is an important nuclear safety issue in case of a severe reactor accident. Namely, the formed very high pressure region induces dynamic loadings on the surrounding structures that may potentially lead to an early release of the radioactive material into the environment. Although the steam explosion events have being studied for several years, the level of the process and consequences understanding is still not adequate. To increase the level of confidence the OECD programme SERENA (Steam Explosion REsolution for Nuclear Applications) was established in 2002. The objectives of the program were to evaluate capabilities of the current generation of the FCI (Fuel-Coolant Interaction) computer codes in predicting the steam explosion induced loads, identifying key FCI phenomena and associated uncertainties impacting the predictability of the steam explosion energetics in the reactor situations and proposing confirmatory research to reduce the uncertainties to acceptable levels for the steam explosion risk assessment. To get a better insight into the most challenging ex-vessel steam explosions, analyses for different locations of the melt release, the cavity water sub-cooling, the primary system pressure overpressure and the triggering time were preformed for a typical pressurized water reactor cavity. The results of some scenarios revealed that significantly higher pressure loads are predicted than obtained in the OECD programme SERENA Phase 1. Among the performed analyses for the central melt pour scenarios, the maximum pressure loads were

Nuclear reactor

International Nuclear Information System (INIS)

Tilliette, Z.

1975-01-01

A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

Secondary coolant circuit for liquid-metal cooled reactor and steam generator for such a circuit

International Nuclear Information System (INIS)

Brachet, A.; Figuet, J.; Guidez, J.; Lions, N.; Traiteur, R.; Zuber, T.

1984-01-01

An upper buffer tank and downstream buffer tank are disposed inside the steam generators. The downstream briffer tank is annular and it surrounds and communicates with a zone of the steam generator through which the liquid metal flows towards the bottom between the exchange zone and the outlet nozzle. The pressure of the inert gas blanket in the downstream buffer volume is more important than this one in the upper buffer volume. The invention applies to fast neutron nuclear reactor cooled by sodium [fr

Steam-generator tube performance: world experience with water-cooled nuclear power reactors during 1978

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1980-01-01

The performance of steam-generator tubes in water-cooled nuclear power reactors during 1978 is reviewed. Tube failures occurred at 31 of the 86 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The number of tubes plugged has decreased dramatically in 1978 compared to the previous year. This is attributed to the diligent application of techniques developed through in-plant experience and research and development programs over the past several years

Evaluation of acoustic resonance at branch section in main steam line. Part 2. Proposal of method for predicting resonance frequency in steam flow

International Nuclear Information System (INIS)

Uchiyama, Yuta; Morita, Ryo

2012-01-01

Flow-induced acoustic resonances of piping system containing closed side-branches are sometimes encountered in power plants. Acoustic standing waves with large amplitude pressure fluctuation in closed side-branches are excited by the unstable shear layer which separates the mean flow in the main piping from the stagnant fluid in the branch. In U.S. NPP, the steam dryer had been damaged by high cycle fatigue due to acoustic-induced vibration under a power uprating condition. Our previous research developed the method for evaluating the acoustic resonance at the branch sections in actual power plants by using CFD. In the method, sound speed in wet steam is evaluated by its theory on the assumption of homogeneous flow, although it may be different from practical sound speed in wet steam. So, it is necessary to consider and introduce the most suitable model of practical sound speed in wet steam. In addition, we tried to develop simplified prediction method of the amplitude and frequency of pressure fluctuation in wet steam flow. Our previous experimental research clarified that resonance amplitude of fluctuating pressure at the top of the branch in wet steam. However, the resonance frequency in steam condition could not be estimated by using theoretical equation as the end correction in steam condition and sound speed in wet steam is not clarified as same reason as CFD. Therefore, in this study, we tried to evaluate the end correction in each dry and wet steam and sound speed of wet steam from experimental results. As a result, method for predicting resonance frequency by using theoretical equation in each wet and dry steam condition was proposed. (author)

GERDA test facility for pressurized water reactors with straight tube steam generators

International Nuclear Information System (INIS)

Ahrens, G.; Haury, G.; Lahner, K.; Schatz, A.

1983-01-01

A number of large-scale experimental facilities have been constructed and operate in order to experiment on the thermodynamic and thermohydraulic behaviour of nuclear facilities in case of LOCA. Most of them were designed for ''large leak'' accidents, but as ''small leak'' accidents became the focus of interest, such experiments were also carried out. Experiments carried out with this arrangement for PWR-type reactors with straight-tube steam generators are only partially evaluable. BBR and B and W therefore cooperated in the construction of the test facility GERDA, designed for testing reactors of BBR design. It supplied relevant experimental results for the nuclear power plant at Muelheim-Kaerlich. (orig.) [de

Experimental study of steam bubble velocities and dimensions in the draught trunk of the AST-500 reactor simulator

International Nuclear Information System (INIS)

Shanin, V.K.; Drobkov, V.P.; Kulakov, I.V.; Khalmeh, M.V.

1988-01-01

Local characteristics for two-phase steam water flow in the vertical channel with 0.45 m diameter and 2 m length, which is the draught trunk of the AST-500 reactor simulator, are investigated. Steam bubble velocities and dimensions were determined by the time-of-flight method using the twinned conductometric transducers. The data obtained testify to the existance of unstable circulation flows in the trunk peripheral region. These flows effect considerably the steam phase motion in the trunk middle part. At the same time the circulation flows to a lesser degree affect steam bubble motion in the trunk low peripheral part and to the lesser degree affect the steam phase in the axial zone near the outlet from the heating section. So the data obtained confirm the conclusion, made earlier, about steam-water flow acceleration in the draught trunk central part

Fuqing nuclear power of nuclear steam turbine generating unit No.1 at the implementation and feedback

International Nuclear Information System (INIS)

Cao Yuhua; Xiao Bo; He Liu; Huang Min

2014-01-01

The article introduces the Fuqing nuclear power of nuclear steam turbine generating unit no.l purpose, range of experience, experiment preparation, implementation, feedback and response. Turn of nuclear steam turbo-generator set flush, using the main reactor coolant pump and regulator of the heat generated by the electric heating element and the total heat capacity in secondary circuit of reactor coolant system (steam generator secondary side) of saturated steam turbine rushed to 1500 RPM, Fuqing nuclear power of nuclear steam turbine generating unit no.1 implementation of the performance of the inspection of steam turbine and its auxiliary system, through the test problems found in the clean up in time, the nuclear steam sweep turn smooth realization has accumulated experience. At the same time, Fuqing nuclear power of nuclear steam turbine generating unit no.1 at turn is half speed steam turbine generator non-nuclear turn at the first, with its smooth realization of other nuclear power steam turbine generator set in the field of non-nuclear turn play a reference role. (authors)

Maintenance and repair of LMFBR steam generators

International Nuclear Information System (INIS)

Verriere, P.; Alanche, J.; Minguet, J.L.

1984-06-01

After some general remarks on the French fast neutron system, this paper presents the state of the program for the construction of fast reactor in France. Then, the general design of Super Phenix 1 steam generator components is outlined and, the in-service monitoring systems and protective devices with which they are equiped are briefly described. The methods used, in the event of leakage, for leak location, steam generator inspection, steam generator repair and putting the affected loop back into service, are discussed. There are two main lines of research, relating respectively to the means of water leak detection in sodium and the inspection arrangements that will be used either periodically, or following a sodium-water reaction. Finally, after a brief description of the steam generator, this paper describes the four incidents (leaks) that occurred on the Phenix steam generator in the course of 1982 and 1983, and the subsequent repair operations

Pressurized water reactor systems

International Nuclear Information System (INIS)

Meyer, P.J.

1975-01-01

Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

Modelling of steam condensation in the primary flow channel of a gas-heated steam generator

International Nuclear Information System (INIS)

Kawamura, H.; Meister, G.

1982-10-01

A new simulation code has been developed for the analysis of steam ingress accidents in high temperatures reactors which evaluates the heat transfer in a steam generator headed by a mixture of helium and water steam. Special emphasis is laid on the analysis of steam condensation in the primary circuit of the steam generator. The code takes wall and bulk condensation into account. A new method is proposed to describe the entrainment of water droplets in the primary gas flow. Some typical results are given. Steam condensation in the primary channel may have a significant effect on temperature distributions. The effect on the heat transferred by the steam generator, however, is found to be not so prominent as might be expected. The reason is discussed. A simplified code will also be described, which gives results with reasonable accuracy within much shorter execution times. This code may be used as a program module in a program simulating the total primary circuit of a high temperature reactor. (orig.) [de

Boiler systems for nuclear powered reactors

International Nuclear Information System (INIS)

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

1979-01-01

A power generating plant which comprises a heat source, at least one main steam turbine and at least one main boiler heated by heat from the heat source and providing the steam to drive the turbine, comprises additionally at least one further steam turbine, smaller than the main turbine, and at least one further boiler, of lower capacity than the main boiler, and heated from the same heat source and providing steam for the further turbine. Particularly advantageous in nuclear power stations, where the heat source is a nuclear reactor, the invention enables peak loads, above the normal continuous rating of the main generators driven by the main turbines, to be met by the further turbine(s) and one or more further generators driven thereby. This enables the main turbines to be freed from the thermal stresses of rapid load changes, which stresses are more easily accommodated by the smaller and thus more tolerant further turbine(s). Thus auxiliary diesel-driven or other independent power plant may be made partly or wholly unnecessary. Further, low-load running which would be inefficient if achieved by means of the main turbine(s), can be more efficiently effected by shutting them down and using the smaller further turbine(s) instead. These latter may also be used to provide independent power for servicing the generating plant during normal operation or during emergency or other shutdown, and in this latter case may also serve as a heat sink for the shutdown reactor

Operation of a steam hydro-gasifier in a fluidized bed reactor

OpenAIRE

Park, Chan Seung; Norbeck, Joseph N.

2008-01-01

Carbonaceous material, which can comprise municipal waste, biomass, wood, coal, or a natural or synthetic polymer, is converted to a stream of methane and carbon monoxide rich gas by heating the carbonaceous material in a fluidized bed reactor using hydrogen, as fluidizing medium, and using steam, under reducing conditions at a temperature and pressure sufficient to generate a stream of methane and carbon monoxide rich gas but at a temperature low enough and/or at a pressure high enough to en...

Hydrogen production by methanol steam reforming carried out in membrane reactor on Cu/Zn/Mg-based catalyst

NARCIS (Netherlands)

Basile, A.; Parmaliana, A.; Tosti, S.; Iulianelli, A.; Gallucci, F.; Espro, C.; Spooren, J.

2008-01-01

The methanol steam reforming (MSR) reaction was studied by using both a dense Pd-Ag membrane reactor (MR) and a fixed bed reactor (FBR). Both the FBR and the MR were packed with a new catalyst based on CuOAl2O3ZnOMgO, having an upper temperature limit of around 350 °C. A constant sweep gas flow rate

BWR type reactors

International Nuclear Information System (INIS)

Tsunoyama, Shigeaki; Tanabe, Akira.

1979-01-01

Purpose: To provide a main steam pressure shock absorber for reflecting the effect of the pressure propagation to coolants surface in the reactor core. Constitution: An annular shock absorber having near the water level through holes for water level measurement is provided to the gap between the skirt of a steam separator and a pressure vessel. Pressure waves are made the rapid closure of a main steam check valve. If arrived from the dome to the shock absorber, are mostly reflected to the side of the dome and give no substantial effects on the water surface. If the through holes are made small enough, the effects of pressure waves passing through the holes are negligible if they reach the water surface. (Kawakami, Y.)

Welded joints engineering design of the primary circuit, surge line and main steam piping of the Angra 2 reactor

International Nuclear Information System (INIS)

Volta, Angelo Roberto; Couto, Jose Gonzalo Villaverde

1995-01-01

The erection of nuclear systems of a Nuclear Power Station is under international requests, that results in a detailed elaboration of documents for the performance of welds. NUCLEN as an engineering design company, responsible for the erection of Angra 2, developed a suitable software program for the elaboration of welding procedure qualifications, tests and examination sequence plans and heat treatment plans applied to primary circuit, surgeline and main steam piping. The paper shows the employed methodology for the elaboration of these documents, as well as the requested engineering design of welding technology and testability in order to assure the stipulated quality level, according to requirements of the specifications, codes and norms. (author). 6 refs

Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

International Nuclear Information System (INIS)

Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

2012-01-01

700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

Core cooling system for reactor

International Nuclear Information System (INIS)

Kondo, Ryoichi; Amada, Tatsuo.

1976-01-01

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

Steam turbines for PWR stations

International Nuclear Information System (INIS)

Muscroft, J.

1989-01-01

The thermodynamic cycle requirements and mechanical design features applying to modern GEC 3000 rev/min steam turbines for pressurised water reactor power stations are reviewed. The most recent developments include machines of 630 MW and 985 MW output which are currently under construction. The importance of service experience with nuclear wet steam turbines associated with a variety of types of water cooled reactor and its relevance to the design of modern 3000 rev/min turbines for pressurised water reactor applications is emphasised. (author)

Fourth international seminar on horizontal steam generators

Energy Technology Data Exchange (ETDEWEB)

Tuomisto, H. [ed.] [IVO Group, Vantaa (Finland); Purhonen, H. [ed.] [VTT, Espoo (Finland); Kouhia, V. [ed.] [Lappeenranta Univ. of Technology (Finland)

1997-12-31

The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries.

Coupled simulation of steam line break accident

International Nuclear Information System (INIS)

Royer, E.; Raimond, E.; Caruge, D.

2000-01-01

The steam line break is a PWR type reactor design accident, which concerns coupled physical phenomena. To control these problems simulation are needed to define and validate the operating procedures. The benchmark OECD PWR MSLB (Main Steam Line Break) has been proposed by the OECD to validate the feasibility and the contribution of the multi-dimensional tools in the simulation of the core transients. First the benchmark OECD PWR MSLB is presented. Then the analysis of the three exercises (system with pinpoint kinetic, three-dimensional core and whole system with three-dimensional core) are discussed. (A.L.B.)

Fourth international seminar on horizontal steam generators

Energy Technology Data Exchange (ETDEWEB)

Tuomisto, H [ed.; IVO Group, Vantaa (Finland); Purhonen, H [ed.; VTT, Espoo (Finland); Kouhia, V [ed.; Lappeenranta Univ. of Technology (Finland)

1998-12-31

The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries.

Fourth international seminar on horizontal steam generators

International Nuclear Information System (INIS)

Tuomisto, H.; Purhonen, H.; Kouhia, V.

1997-01-01

The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries

Steam plant for pressurized water reactors

International Nuclear Information System (INIS)

Anon.

1983-01-01

This book discusses the research and development organisations and users to highlight those aspects of the steam turbine and associated plant which are particularly related to the PWR system. The contents include: Characteristics of the steam system (including feed train, dump system and safety aspects); overall design aspects of high and half speed turbines; design aspects of the steam generator and seismic considerations; moisture separators and reheaters; feed pumps and their drives; water treatment; safety related valves; operational experience; availability and performance

Hot steam header of a high temperature reactor as a benchmark problem

International Nuclear Information System (INIS)

Demierre, J.

1990-01-01

The International Atomic Energy Agency (IAEA) initiated a Coordinated Research Programme (CRP) on ''Design Codes for Gas-Cooled Reactor Components''. The specialists proposed to start with a benchmark design of a hot steam header in order to get a better understanding of the methods in the participating countries. The contribution of Switzerland carried out by Sulzer. The following report summarized the detailed calculations of dimensioning procedure and analysis. (author). 5 refs, 2 figs, 2 tabs

Water box for steam generator

International Nuclear Information System (INIS)

Lecomte, Robert; Viaud, Michel.

1975-01-01

This invention relates to a water box for connecting an assembly composed of a vertical steam generator and a vertical pump to the vessel of the nuclear reactor, the assembly forming the primary cooling system of a pressurised water reactor. This invention makes it easy to dismantle the pump on the water box without significant loss of water in the primary cooling system of the reactor and particularly without it being necessary to drain the water contained in the steam generator beforehand. It makes it possible to shorten the time required for dismantling the primary pump in order to service or repair it and makes dismantling safer in that the dismantling does not involve draining the steam generator and therefore the critical storage of a large amount of cooling water that has been in contact with the fuel assemblies of the nuclear reactor core [fr

Assessment of weld joints of steam generator of prototype fast breeder reactor by microfocal radiography

International Nuclear Information System (INIS)

Venkatraman, B.; Saravanan, T.; Jayakumar, T.; Kalyanasundaram, P.; Raj, B.

2004-01-01

The tube to tubesheet (TTS) welds of steam generator of Prototype Fast Breeder Reactor (PFBR) are quite critical. Sodium flows on shell side and water on tube side. Any failure would thus be catastrophic. Apart from defects such as porosities, wall thinning due to concavity is endemic in such joints and needs to be detected. This paper presents the methodologies developed for quantitative evaluation of defects including wall thinning due to concavity in the TTS welds by micro focal radiography. The method has been successfully adopted in the shop floor for the evaluation of TTS welds of steam generator and evaporator. (author)

Steam generator materials and secondary side water chemistry in nuclear power stations

International Nuclear Information System (INIS)

Rudelli, M.D.

1979-04-01

The main purpose of this work is to summarize the European and North American experiences regarding the materials used for the construction of the steam generators and their relative corrosion resistance considering the water chemestry control method. Reasons underlying decision for the adoption of Incoloy 800 as the material for the secondary steam generator system for Atucha I Nuclear Power Plant (Atucha Reactor) and Embalse de Rio III Nuclear Power Plant (Cordoba Reactor) are pointed out. Backup information taken into consideration for the decision of utilizing the All Volatil Treatment for the water chemistry control of the Cordoba Reactor is detailed. Also all the reasonswhich justify to continue with the congruent fosfatic method for the Atucha Reactor are analyzed. Some investigation objectives which would eventually permit the revision of the decisions taken on these subjects are proposed. (E.A.C.) [es

Twin header bore welded steam generator for pressurized water reactors

International Nuclear Information System (INIS)

Davies, R.J.; Hirst, B.

1979-01-01

A description is given of a pressurized water reactor (PWR) steam generator concept, several examples of which have been in service for up to fourteen years. Details are given of the highly successful service record of this equipment and the features which have been incorporated to minimize corrosion and deposition pockets. The design employs a vertical U tube bundle carried off two horizontal headers to which the tubes are welded by the Foster Wheeler Power Products (FWPP) bore welding process. The factors to be considered in uprating the design to meet the current operating conditions for a 1000 MW unit are discussed. (author)

Research and engineering application of coordinated instrumentation control and protection technology between reactor and steam turbine generator on nuclear power plant

International Nuclear Information System (INIS)

Sun Xingdong

2014-01-01

The coordinated instrumentation control and protection technology between reactor and steam turbine generator (TG) usually is very significant and complicated for a new construction of nuclear power plant, because it carries the safety, economy and availability of nuclear power plant. Based on successful practice of a nuclear power plant, the experience on interface design and hardware architecture of coordinated instrumentation control and protection technology between reactor and steam turbine generator was abstracted and researched. In this paper, the key points and engineering experience were introduced to give the helpful instructions for the new project. (author)

Report on US-Japan 1983 meetings on steam generators

International Nuclear Information System (INIS)

1984-04-01

This is a report on a trip to Japan by personnel of the US Nuclear Regulatory Commission in 1983 to exchange information on steam generators of nuclear power plants. Steam generators of Japanese pressurized water reactors have experienced nearly all of the forms of degradation that have been experienced in US recirculating-type steam generators, except for denting and pitting. More tubes have been plugged per year of reactor operation in Japanese than in US steam generators, but much of the Japanese tube plugging is preventative rather than the result of leaks experienced. The number of leaks per reactor year is much smaller for Japanese than for US steam generators. No steam generators have been replaced in Japan while several have been replaced in the US. The Japanese experience may be related to their very stringent inspection and maintenance programs for steam generators

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)

Steam-generator tube failures: world experience in water-cooled nuclear power reactors in 1974

International Nuclear Information System (INIS)

Hare, M.G.

1976-01-01

Steam-generator tube failures were reported at 25 of 59 water-cooled nuclear power reactors surveyed in 1974, compared to 11 of 49 in 1973. A summary is presented of these failures, most of which, where the cause is known, were the result of corrosion. Water chemistry control, inspection and repair procedures, and failure rates are discussed

Thermal hydraulic studies in steam generator test facility

International Nuclear Information System (INIS)

Vinod, V.; Suresh Kumar, V.A.; Noushad, I.B.; Ellappan, T.R.; Rajan, K.K.; Rajan, M.; Vaidyanathan, G.

2005-01-01

Full text of publication follows: A 500 MWe fast breeder reactor is being constructed at Kalpakkam, India. This is a sodium cooled reactor with two primary and two secondary sodium loops with total 8 steam generators. The typical advantage of fast breeder plants is the high operating temperature of steam cycles and the high plant efficiency. To produce this high pressure and high temperature steam, once through straight tube vertical sodium heated steam generators are used. The steam is generated from the heat produced in the reactor core and being transported through primary and secondary sodium circuits. The steam generator is a 25 m high middle supported steam generator with expansion bend and 23 m heat transfer length. Steam Generator Test Facility (SGTF) constructed at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam aims at performing various tests on a 5.5 MWt steam generator. This vertically simulated test article is similar in all respects to the proposed 157 MWt steam generator module for the Prototype Fast Breeder Reactor (PFBR), with reduced number of tubes. Heat transfer performance tests are done with this 19 tube steam generator at various load conditions. Sodium circuit for the SGTF is equipped with oil fired heater as heat source and centrifugal sodium pump, to pump sodium at 105 m 3 /hr flow rate. Other typical components like sodium to air heat exchanger, sodium purification system and hydrogen leak detection system is also present in the sodium circuit. High pressure steam produced in the steam generator is dumped in a condenser and recycled. Important tests planned in SGTF are the heat transfer performance test, stability test, endurance test and performance test of steam generator under various transients. The controlled operation of steam generator will be studied with possible control schemes. A steady state simulation of the steam generator is done with a mathematical model. This paper gives the details of heat transfer

Sodium/water reactions in steam generators of liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Hori, M.

1980-01-01

The status of the research and development on sodium/water reactions resulting from the leakage of water into sodium in LMFBR steam generators is reviewed. The importance of sodium/water reaction phenomena in the design and operation of steam generators is discussed. The effects of sodium/water reactions are evaluated and methods of protection against these phenomena are surveyed. The products of chemical reactions between sodium and water under steam generator conditions are H 2 , NaOH, Na 2 O and NaH. Together with the temperature rise due to the associated exothermic reaction, these reaction products cause effects such as self-wastage, single- and multi-target wastage, and rapid pressure increase, depending on the size of the leak hole or the magnitude of leak rate. As for the wastage phenomena of small leaks, the effects of various factors have been studied and experimental correlations, as well as some theoretical work, have been performed. To investigate the pressure phenomena of a large leak, large-scale tests have been conducted by various organizations, and the computer codes to analyse these phenomena have been developed and verified by experiments. In the design of steam generators, an initial failure up to a hypothetical double-ended guillotine rupture of a single heat transfer tube is widely used as the design basis leak. Protection systems for LMFBR plants consist of leak detection devices, leak termination devices, and reaction pressure relief devices. From analyses based on research and development activities, as well as from experience with leaks in steam generator test loops and reactor plants, it has been confirmed that protection systems can satisfactorily be designed to accommodate leak incidents in LMFBR plants. (author)

Analysis and modeling of flow-blockage-induced steam explosion events in the high-flux isotope reactor

International Nuclear Information System (INIS)

Taleyarkhan, R.P.; Georgevich, V.; Nestor, C.W.; Gat, U.; Lepard, B.L.; Cook, D.H.; Freels, J.; Chang, S.J.; Luttrell, C.; Gwaltney, R.C.

1994-01-01

This article provides a perspective overview of the analysis and modeling work done to evaluate the threat from steam explosion loads in the High-Flux Isotope Reactor (HFIR) during flow blockage events. The overall work scope included modeling and analysis of core-melt initiation, melt propagation, bounding and best-estimate steam explosion energetics, vessel failure from fracture, bolts failure from exceedance of elastic limits, and, finally, missile evolution and transport. Aluminum ignition was neglected. Evaluations indicated that a thermally driven steam explosion with more than 65 MJ of energy insertion in the core region over several milliseconds would be needed to cause a sufficiently energetic missile with a capacity to cause early confinement failure. This amounts to about 65% of the HFIR core mass melting and participating in a steam explosion. Conservative melt propagation analyses have indicated that at most only 24% of the HFIR core mass could melt during flow blockage events under full-power conditions. 19 refs., 11 figs

A study on the evaluation of vibration effect and the development of vibration reduction method for Wolsung unit 1 main steam piping

Energy Technology Data Exchange (ETDEWEB)

Lee, Hyun; Kim, Yeon Whan [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Kim, Tae Ryong; Park, Jin Ho [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

1996-08-01

The main steam piping of nuclear power plant which runs between steam generator and high pressure turbine has been experienced to have a severe effect on the safe operation of the plant due to the vibration induced by the steam flowing inside the piping. The imposed cyclic loads by the vibration could result in the degradation of the related structures such as connection parts between main instruments, valves, pipe supports and building. The objective of the study is to reduce the vibration level of Wolsung nuclear power plant unit 1 main steam pipeline by analyzing vibration characteristics of the piping, identifying sources of the vibration and developing a vibration reduction method .The location of the maximum vibration is piping between the main steam header and steam chest .The stress level was found to be within the allowable limit .The main vibration frequency was found to be 4{approx}6 Hz which is the same as the natural frequency from model test .A vibration reduction method using pipe supports of energy absorbing type(WEAR)is selected .The measured vibration level after WEAR installation was reduced about 36{approx}77% in displacement unit (author). 36 refs., 188 figs.

Experimental investigation of non-condensable gases effect on operation of VVER steam generator in condensation mode

International Nuclear Information System (INIS)

Efanov, A. D.; Kalyakin, S. G.; Morozov, A. V.; Remizov, O. V.; Tsyganok, A. A.; Generalov, V. N.; Berkovich, V. M.; Taranov, G. S.

2008-01-01

To provide the safety in new Russian NPP designs, protection passive systems which don't depend upon human errors are widely used. In terms of safety, the design of NPP of new generation (NPP-2006) falls into the class of advanced NPPs. In the event of an beyond design basis accident with the rupture of the reactor primary circuit and accompanied by the loss of ac sources, the use of passive safety systems are provided for necessary core cooling. Among these is passive heat removal system (PHRS). In the case of leakage in the primary circuit this system ensures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam coming to SG piping from the reactor. As a result, the condensate from steam generators arrives at the core providing its additional cooling. The SG condensation capacity can be adversely affected by the presence of non-condensable gases in the primary circuit of the reactor. Their main sources are nitrogen arriving at the circuit, as hydro accumulators actuate, products of radiolysis of water and air drawn in from the containment through the pipeline rupture. The accumulation of non-condensable gases in SG piping can result in degradation of its condensation capacity to the extent that condensation completely terminates. In this case, the core cooling conditions may be impaired. To experimental investigation of the condensation mode of operation of WER steam generator, a large scale HA2M-SG test rig was constructed at the SSC RF IPPE. The test rig incorporates: buffer tank, equipped by steam supply system; SG model with volumetric-power scale is 1:46; PHRS heat exchanger imitator, cooling by process water. The rig main equipment connected by pipelines and equipped by valves. The elevations of the main equipment correspond to those of reactor project. The rig maximum operating parameters: steam pressure - 1.6 MPa, temperature - 200 Celsius degrees. Experiments at the HA2M-SG test rig have been

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

Energy Technology Data Exchange (ETDEWEB)

Brown, C.S., E-mail: csbrown3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Raleigh, NC 27695-7909 (United States); Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3870 (United States); Kucukboyaci, V., E-mail: kucukbvn@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Sung, Y., E-mail: sungy@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2016-12-01

Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

International Nuclear Information System (INIS)

Brown, C.S.; Zhang, H.; Kucukboyaci, V.; Sung, Y.

2016-01-01

Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

Main boiler feed pump for fast breeder test reactor. Failure analysis and remedial measures

International Nuclear Information System (INIS)

Iyer, M.A.K.; Chande, S.K.; Raghuvir, A.D.; Baskar, S.; Kale, R.D.

1994-01-01

A small capacity ten stage 670 kw feed water pump is used for supplying feed water at a temperature of 190 deg C to a once through steam generator in the Fast Breeder Test Reactor at Kalpakkam. During preparatory heating up stage to commission the steam generator the pump suffered a severe loss of suction which resulted in failure of hydrostatic journal bearings and extensive damage to pump internals. This paper discusses the detailed mechanism of loss of suction, details of damage to the pump and various modifications carried out to prevent recurrence of the problem. (author). 4 refs., 3 figs., 2 tabs

SCW Pressure-Channel Nuclear Reactors: Some Design Features and Concepts

International Nuclear Information System (INIS)

Duffey, R.B.; Pioro, I.L.; Gabaraev, B.A.; Kuznetsov, Yu. N.

2006-01-01

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950's and 1960's in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with supercritical water (SCW) became attractive again as the ultimate development path for water-cooling. The main objectives of using SCW in nuclear reactors are 1) to increase the thermal efficiency of modern nuclear power plants (NPPs) from 33 -- 35% to about 40 -- 45%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (∼$ 1000 US/kW). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625 deg. C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia. Design features related to both channels and fuel bundles are discussed in this paper. Also, Russian experience with operating supercritical steam heaters at NPP is presented. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal energy systems. (authors)

Remote-controlled television for locating leaking tubes in pressurized-water reactor steam generators

International Nuclear Information System (INIS)

Cormault, P.; Denis, J.

1978-01-01

The Scarabee system is designed for observation of the tubes in water boxes of pressurized-water reactor nuclear-power-station steam generators. It consists essentially of a camera and a projector used as a marker, both of which swivel freely. The whole unit is housed in a water-tight container which can easily be decontaminated. Remote control of camera and marker movement is carried out from a console. (author)

ORTAP: a nuclear steam supply system simulation for the dynamic analysis of high temperature gas cooled reactor transients

International Nuclear Information System (INIS)

Cleveland, J.C.; Hedrick, R.A.; Ball, S.J.; Delene, J.G.

1977-01-01

ORTAP was developed to predict the dynamic behavior of the high temperature gas cooled reactor (HTGR) Nuclear Steam Supply System for normal operational transients and postulated accident conditions. It was developed for the Nuclear Regulatory Commission (NRC) as an independent means of obtaining conservative predictions of the transient response of HTGRs over a wide range of conditions. The approach has been to build sufficient detail into the component models so that the coupling between the primary and secondary systems can be accurately represented and so that transients which cover a wide range of conditions can be simulated. System components which are modeled in ORTAP include the reactor core, a typical reheater and steam generator module, a typical helium circulator and circulator turbine and the turbine generator plant. The major plant control systems are also modeled. Normal operational transients which can be analyzed with ORTAP include reactor start-up and shutdown, normal and rapid load changes. Upset transients which can be analyzed with ORTAP include reactor trip, turbine trip and sudden reduction in feedwater flow. ORTAP has also been used to predict plant response to emergency or faulted conditions such as primary system depressurization, loss of primary coolant flow and uncontrolled removal of control poison from the reactor core

Trip report: United States LMFBR Steam Generator Team. IAEA symposium, Bensberg, Germany, October 14--17, 1974

International Nuclear Information System (INIS)

1974-01-01

Information is presented concerning steam generator design characteristics for the AFR reactor, SNR reactor, PHENIX reactor, SUPER PHENIX reactor, MONJU reactor, and BN-350 reactor; steam generator development programs for West Germany, France, Japan, U. K., and the U. S. S. R.; and the fabrication and inspection of steam generator components. Steam generator performance and maintenance requirements for operating LMFBR reactors are reviewed. (U.S.)

Steam generator for pressurized-water reactors

International Nuclear Information System (INIS)

Michel, E.

1971-01-01

In the steam generator for a PWR the central fall space of a U-tube bundel heat exchanger is used as a preliminary cyclon separator. The steam escaping upwards, which is largely free of water, can flow through the residual heating surface, i.e. the U-tube turns. In this way substantial drying and less superheating by the heat still added becomes possible. In its upper part the central fall space for the water separated in the preliminary separator, enclosed by a cylindrical guide wall and the U-tube bundle, is provided with tangential inlet slots. Through these, the water-steam mixture steams out of the section of the vertical legs of the U-tube bundle into the fall space. Above the inlet slots the rising space is closed by means of a turn-round plate. At the lower end of the guide wall outlet, slots are provided for the water flowing downwards and radially outwards into the unfilled space. (DG/PB) [de

Draining down of a nuclear steam generating system

International Nuclear Information System (INIS)

Jawor, J.C.

1987-01-01

The method is described of draining down contained reactor-coolant water from the inverted vertical U-tubes of a vertical-type steam generator in which the upper, inverted U-shaped ends of the tubes are closed and the lower ends thereof are open. The steam generator is part of a nuclear powered steam generating system wherein the reactor coolant water is normally circulated from and back into the reactor via a loop comprising the steam generator and inlet and outlet conduits connected to the lower end of the steam generator. The method comprises continuously introducing a gas which is inert to the system and which is under pressure above atmospheric pressure into at least one of the downwardly facing open ends of each of the U-tubes from below the tube sheet in which the open ends of the U-tubes are mounted adjacent the lower end of the steam generator, while permitting the water to flow out from the open ends of the U-tubes

Steam Generator Tube Integrity Program: Surry Steam Generator Project, Hanford site, Richland, Benton County, Washington: Environmental assessment

International Nuclear Information System (INIS)

1980-03-01

The US Nuclear Regulatory Commission (NRC) has placed a Nuclear Regulatory Research Order with the Richland Operations Office of the US Department of Energy (DOE) for expanded investigations at the DOE Pacific Northwest Laboratory (PNL) related to defective pressurized water reactor (PWR) steam generator tubing. This program, the Steam Generator Tube Integrity (SGTI) program, is sponsored by the Metallurgy and Materials Research Branch of the NRC Division of Reactor Safety Research. This research and testing program includes an additional task requiring extensive investigation of a degraded, out-of-service steam generator from a commercial nuclear power plant. This comprehensive testing program on an out-of-service generator will provide NRC with timely and valuable information related to pressurized water reactor primary system integrity and degradation with time. This report presents the environmental assessment of the removal, transport, and testing of the steam generator along with decontamination/decommissioning plans

CO-free hydrogen production by ethanol steam reforming in a Pd-Ag membrane reactor

NARCIS (Netherlands)

Basile, A.; Gallucci, F.; Iulianelli, A.; Tosti, S.

2008-01-01

In this work, the ethanol steam reforming (ESR) reaction has been studied by using a dense Pd-Ag membrane reactor (MR) by varying the water/ethanol molar ratio between 3:1 and 9:1 in a temperature range of 300-400°C and at 1.3 bar as reaction pressure. The MR was packed with a commercial Ru-based

Nuclear reactor monitoring device

International Nuclear Information System (INIS)

Mihashi, Ishi; Honma, Hitoshi.

1993-01-01

The monitoring device of the present invention comprises a reactor core/reactor system data measuring and controlling device, a radioactivity concentration calculation device for activated coolants for calculating a radioactivity concentration of activated coolants in a main steam and reactor water by using an appropriate physical model, a radioactivity concentration correlation and comparison device for activated coolants for comparing correlationship with a radiation dose and an abnormality alarm device. Since radioactivity of activated primary coolants is monitored at each of positions in the reactor system and occurrence of leakage and the amount thereof from a primary circuit to a secondary circuit is monitored if the reactor has secondary circuit, integrity of the reactor system can be ensured and an abnormality can be detected rapidly. Further, radioactivity concentration of activated primary circuit coolants, represented by 16 N or 15 C, is always monitored at each of positions of PWR primary circuits. When a heat transfer pipe is ruptured in a steam generator, leakage of primary circuit coolants is detected rapidly, as well as the amount of the leakage can be informed. (N.H.)

Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 2. TRAB-3D/SMABRE Calculation of the OECD/ NRC PWR MSLB Benchmark

International Nuclear Information System (INIS)

Daavittila, A.; Haemaelaeinen, A.; Kyrki-Rajamaki, R.

2001-01-01

All three exercises of the OECD/NRC Pressurized Water Reactor (PWR) Main-Steam-Line-Break (MSLB) Benchmark were calculated at VTT Energy. The SMABRE thermal-hydraulics code was used for the first exercise, the plant simulation with point-kinetics neutronics. The second exercise was calculated with the TRAB-3D three-dimensional reactor dynamics code. The third exercise was calculated with the combination TRAB-3D/SMABRE. Both codes have been developed at VTT Energy. The results of all the exercises agree reasonably well with those of the other participants; thus, instead of reporting the results, this paper concentrates on describing the computational aspects of the calculation with the foregoing codes and on some observations of the sensitivity of the results. In the TRAB-3D neutron kinetics, the two-group diffusion equations are solved in homogenized fuel assembly geometry with an efficient two-level nodal method. The point of the two-level iteration scheme is that only one unknown variable per node, the average neutron flux, is calculated during the inner iteration. The nodal flux shapes and cross sections are recalculated only once in the outer iteration loop. The TRAB-3D core model includes also parallel one-dimensional channel hydraulics with detailed fuel models. Advanced implicit time discretization methods are used in all submodels. SMABRE is a fast-running five-equation model completed by a drift-flux model, with a time discretization based on a non-iterative semi-implicit algorithm. For the third exercise of the benchmark, the TMI-1 models of TRAB-3D and SMABRE were coupled. This was the first time these codes were coupled together. However, similar coupling of the HEXTRAN and SMABRE codes has been shown to be stable and efficient, when used in safety analyses of Finnish and foreign VVER-type reactors. The coupling used between the two codes is called a parallel coupling. SMABRE solves the thermal hydraulics both in the cooling circuit and in the core

On the potential of nickel catalysts for steam reforming in membrane reactors

Energy Technology Data Exchange (ETDEWEB)

Pieterse, J.A.Z.; Boon, J.; Van Delft, Y.C.; Dijkstra, J.W.; Van den Brink, R.W. [Energy research Center of the Netherlands, P.O. Box 1, 1755 ZG Petten (Netherlands)

2010-10-15

Hydrogen membrane reactors have been identified as a promising option for hydrogen production for power generation from natural gas with pre-combustion decarbonisation. While Pd or Pd-alloy membranes already provide good hydrogen permeances the most suitable catalyst design for steam reforming in membrane reactors (SRMR) is yet to be identified. This contribution aims to provide insight in the suitability of nickel based catalysts in SRMR. The use of nickel (Ni) catalysts would benefit the cost-effectiveness of membrane reactors and therefore its feasibility. For this, the activity of nickel catalysts in SRMR was assessed with kinetics reported in literature. A 1D model was composed in order to compare the hydrogen production rates derived from the kinetics with the rate of hydrogen withdrawal by permeation. Catalyst stability was studied by exposing the catalysts to reformate gas with two different H/C ratios to mimic the hydrogen lean reformate gas in the membrane reactor. For both the activity (modeling) and stability study the Ni-based catalysts were compared to relevant catalyst compositions based on rhodium (Rh). Using the high pressure kinetics reported for Al2O3 supported Rh and MgAl2O4 and Al2O3 supported Ni catalyst it showed that Ni and Rh catalysts may very well provide similar hydrogen production rates. Interestingly, the stability of Ni-based catalysts proved to be superior to precious metal based catalysts under exposure to simulated reformate feed gas with low H/C molar ratio. A commercial (pre-)reforming Ni-based catalyst was selected for further testing in an experimental membrane reactor for steam reforming at high pressure. During the test period 98% conversion at 873 K could be achieved. The conversion was adjusted to approximately 90% and stable conversion was obtained during the test period of another 3 weeks. Nonetheless, carbon quantification tests of the Ni catalyst indicated that a small amount of carbon had deposited onto the catalyst

Main Steam Line Break Mass/Energy and Pressure/Temperature Analysis for the Environmental Qualification

International Nuclear Information System (INIS)

Park, Yong-Chan; Song, Dong-Soo; Jun, Hwang-Yong

2006-01-01

The Main steam line break(MSLB) occurring inside a reactor containment structure may result in significant releases of high energy fluid to the containment, possibly result in high containment pressure and temperature. The MSLB accident, along with the Loss Of Coolant Accident, is a design basis accident for determining the peak containment pressure and temperature. The analysis for a MSLB for inside containment should be performed to justify the structural integrity and equipment qualification in accordance with revision 1 of Reg. Guide 1.89. Rev1(1984), which is also required as part of obtaining the extended operating license for WestingHouse(WH) 3-Loops Nuclear Power Plant(NPP). Now, the WH NPP has been performed power uprating. Therefore, all initial conditions, setpoints and uncertainties were considered with MSLB analysis for environment qualification(EQ). The transient was analyzed to determine the worst set of mass and energy releases that impact the EQ aspects of safety related equipment inside containment. The most limiting single failure in this event was determined by a sensitivity study. The MSLB event was analyzed for a full set of power conditions and break sizes

Analysis and modeling of flow blockage-induced steam explosion events in the High-Flux Isotope Reactor

International Nuclear Information System (INIS)

Taleyarkhan, R.P.; Georgevich, V.; Lestor, C.W.; Gat, U.; Lepard, B.L.; Cook, D.H.; Freels, J.; Chang, S.J.; Luttrell, C.; Gwaltney, R.C.; Kirkpatrick, J.

1993-01-01

This paper provides a perspective overview of the analysis and modeling work done to evaluate the threat from steam explosion loads in the High-Flux Isotope Reactor during flow blockage events. The overall workscope included modeling and analysis of core melt initiation, melt propagation, bounding and best-estimate steam explosion energetics, vessel failure from fracture, bolts failure from exceedance of elastic limits, and finally, missile evolution and transport. Aluminum ignition was neglected. Evaluations indicated that a thermally driven steam explosion with more than 65 MJ of energy insertion in the core region over several miliseconds would be needed to cause a sufficiently energetic missile with a capacity to cause early confinement failure. This amounts to about 65% of the HFIR core mass melting and participating in a steam explosion. Conservative melt propagation analyses have indicated that at most only 24% of the HFIR core mass could melt during flow blockage events under full-power conditions. Therefore, it is judged that the HFIR vessel and top head structure will be able to withstand loads generated from thermally driven steam explosions initiated by any credible flow blockage event. A substantial margin to safety was demonstrated

Advanced Catalysis Technologies: Lanthanum Cerium Manganese Hexaaluminate Combustion Catalysts for Flat Plate Reactor for Compact Steam Reformers

Science.gov (United States)

2008-12-01

packed-bed steam reformer reactor using an open-flame or radiant burner as the heat source, the rate of heat transfer is limited by wall film and bed...resistances. Heat transfer can be effectively improved by replacing the burner /packed-bed system with parallel channels containing metal foam...combustion reactor was tested using the hexaaluminate catalyst in pellets and supported on FeCrAlloy metal foam. Both tests burned propane and JP-8

Sorption-enhanced steam methane reforming in fluidized bed reactors

Energy Technology Data Exchange (ETDEWEB)

Johnsen, Kim

2006-10-15

Hydrogen is considered to be an important potential energy carrier; however, its advantages are unlikely to be realized unless efficient means can be found to produce it without generation of CO{sub 2}. Sorption-enhanced steam methane reforming (SE-SMR) represent a novel, energy-efficient hydrogen production route with in situ CO{sub 2} capture, shifting the reforming and water gas shift reactions beyond their conventional thermodynamic limits. The use of fluidized bed reactors for SE-SMR has been investigated. Arctic dolomite, a calcium-based natural sorbent, was chosen as the primary CO{sub 2}-acceptor in this study due to high absorption capacity, relatively high reaction rate and low cost. An experimental investigation was conducted in a bubbling fluidized bed reactor of diameter 0.1 m, which was operated cyclically and batch wise, alternating between reforming/carbonation conditions and higher-temperature calcination conditions. Hydrogen concentrations of >98 mole% on a dry basis were reached at 600 C and 1 atm, for superficial gas velocities in the range of {approx}0.03-0.1 m/s. Multiple reforming-regeneration cycles showed that the hydrogen concentration remained at {approx}98 mole% after four cycles. The total production time was reduced with an increasing number of cycles due to loss of CO{sub 2}-uptake capacity of the dolomite, but the reaction rates of steam reforming and carbonation seemed to be unaffected for the conditions investigated. A modified shrinking core model was applied for deriving carbonation kinetics of Arctic dolomite, using experimental data from a novel thermo gravimetric reactor. An apparent activation energy of 32.6 kj/mole was found from parameter fitting, which is in good agreement with previous reported results. The derived rate expression was able to predict experimental conversion up to {approx}30% very well, whereas the prediction of higher conversion levels was poorer. However, the residence time of sorbent in a continuous

The main trends of work on ageing and lifetime management of NPP

International Nuclear Information System (INIS)

Dragunov, Yu.

1994-01-01

The main trends of work on aging and lifetime management of nuclear power plants(NPP) in Russia are described including Russian research programs on NPP components lifetime assurance, technical assistance program for CIS, measures for life extension of reactor pressure vessels, steam generator collectors and components for advances reactor plants. Figs and tabs

Feedback phenomena in nuclear reactors

International Nuclear Information System (INIS)

Fiebig, R.

1977-01-01

It is investigated what influence the thermodynamic behaviour of the steam dome of a reactor with pressure autocontrol has on the dynamics of the reactor system. For automatic control, either the circuit water must be thermally coupled with the steam dome or, without coupling, there must be a sufficiently large subcooling of the reactor core. The coupling mechanisms between water and steam in the steam dome to be considered are heat conduction, boiling, and condensation. A heat sink in the steam dome enforces a thermodynamic equilibrium between water and steam and provides good autocontrol properties. Without a heat sink, thermal heat coupling is ended when the pressure rises. Nevertheless, with direct contact between circuit and steam dome the reactor remains controllable. At the reactor of the NCS-80, where the circuit is separated from the steam dome by a buffer volume, autocontrol takes place with a heat sink in the steam dome and with sufficient shifting of the working point into the subcooled region caused by the rising of bubbles. (orig.) [de

Limit the effects of secondary circuit water or steam piping breaks in the reactor building

International Nuclear Information System (INIS)

Nachev, N.

2001-01-01

The existing design of the WWER-1000 Model 320 does not include provisions against the local mechanical effects of pipe ruptures of the secondary system piping. This situation may lead to accidental effects beyond the design basis of the plant in case of a postulated secondary pipe rupture event. The aim of the present safety enhancement measure is to overcome this safety deficit, that means to carry out some analyses and to suggest protection measures, by which the specified design basis of the plant concerning secondary circuit design basis accidents will be assured. The systems to be considered include the main steam lines (MSL) and the main feedwater lines (MFWL) in the safety related system areas. These areas are the system portions, which are located in the reactor building (containment and room A820 outside the containment). The pipe rupture effects to be considered include the local effects, that means pipe whip impact and jet forces on the adjacent equipment and structures, as well as reaction forces due to blowdown thrust forces and pressure waves in the broken piping system. (author)

IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Presentations

International Nuclear Information System (INIS)

2011-01-01

concepts incorporating innovative systems and components, as well as advanced fuel and fuel cycle technologies. In particular, innovative heat exchangers and steam generators are key to significanly reduce the capital cost of the NSSS of the fast reactors. The IAEA, within the framework of its Nuclear Energy Department’s Technical Working Group on Fast Reactors (TWG-FR), assists Member States activities in these technology development areas by providing an umbrella for information exchange [topical Technical Meetings (TMs), Workshops and large Conferences] and collaborative R&D [Coordinated Research Projects (CRPs)]. This topical TM is addressing Member States’ expressed information exchange needs in the field of advanced fast reactor design features, with particular attention to innovative heat exchangers and steam generators

Energy and exergy analysis of the turbo-generators and steam turbine for the main feed water pump drive on LNG carrier

International Nuclear Information System (INIS)

Mrzljak, Vedran; Poljak, Igor; Mrakovčić, Tomislav

2017-01-01

Highlights: • Two low-power steam turbines in the LNG carrier propulsion plant were investigated. • Energy and exergy efficiencies of both steam turbines vary between 46% and 62%. • The ambient temperature has a low impact on exergy efficiency of analyzed turbines. • The maximum efficiencies area of both turbines was investigated. • A method for increasing the turbo-generator efficiencies by 1–3% is presented. - Abstract: Nowadays, marine propulsion systems are mainly based on internal combustion diesel engines. Despite this fact, a number of LNG carriers have steam propulsion plants. In such plants, steam turbines are used not only for ship propulsion, but also for electrical power generation and main feed water pump drive. Marine turbo-generators and steam turbine for the main feed water pump drive were investigated on the analyzed LNG carrier with steam propulsion plant. The measurements of various operating parameters were performed and obtained data were used for energy and exergy analysis. All the measurements and calculations were performed during the ship acceleration. The analysis shows that the energy and exergy efficiencies of both analyzed low-power turbines vary between 46% and 62% what is significantly lower in comparison with the high-power steam turbines. The ambient temperature has a low impact on exergy efficiency of analyzed turbines (change in ambient temperature for 10 °C causes less than 1% change in exergy efficiency). The highest exergy efficiencies were achieved at the lowest observed ambient temperature. Also, the highest efficiencies were achieved at 71.5% of maximum developed turbo-generator power while the highest efficiencies of steam turbine for the main feed water pump drive were achieved at maximum turbine developed power. Replacing the existing steam turbine for the main feed water pump drive with an electric motor would increase the turbo-generator energy and exergy efficiencies for at least 1–3% in all analyzed

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

An investigation of decreasing reactor coolant inventory as a mechanism to reduce power during a boiling water reactor anticipated transient without scram

International Nuclear Information System (INIS)

Peterson, C.E.; Chexal, V.K.; Gose, G.C.; Hentzen, R.D.; Layman, W.H.

1985-01-01

Under certain anticipated transient without scram (ATWS) sequences for a boiling water reactor, it would be desirable to reduce system power, particularly where the primary system has been isolated by closure of all main steam isolation valves and is discharging steam through its safety/relief valve system to the suppression pool. Reducing reactor power increases the time available to shut down the reactor by minimizing the heat dumped to the suppression pool and by helping to keep the suppression pool temperature within limits. Under proposed emergency procedure guidelines for the ATWS event, the reactor water level would be lowered to reduce reactor power. The analyses provide an assessment of the power level that would be attained, assuming the reactor operators were to reduce the the downcomer level down to the top of the active fuel

CFD Analysis of Random Turbulent Flow Load in Steam Generator of APR1400 Under Normal Operation Condition

International Nuclear Information System (INIS)

Lim, Sang Gyu; You, Sung Chang; Kim, Han Gon

2011-01-01

Regulatory guide 1.20 revision 3 of the Nuclear Regulatory Committee (NRC) modifies guidance for vibration assessments of reactor internals and steam generator internals. The new guidance requires applicants to provide a preliminary analysis and evaluation of the design and performance of a facility, including the safety margins of during normal operation and transient conditions anticipated during the life of the facility. Especially, revision 3 require rigorous assessments of adverse flow effects in the steam dryer cased by flow-excited acoustic and structural resonances such as the abnormality from power-uprated BWR cases. For two nearly identical nuclear power plants, the steam system of one BWR plant experienced failure of steam dryers and the main steam system components when steam flow was increased by 16 percent for extended power uprate (EPU). The mechanisms of those failures have revealed that a small adverse flow changing from the prototype condition induced severe flow-excited acoustic and structural resonances, leading to structural failures. In accordance with the historical background, therefore, potential adverse flow effects should be evaluated rigorously for steam generator internals in both BWR and Pressurized Water Reactor (PWR). The Advanced Power Reactor 1400 (APR1400), an evolutionary light water reactor, increased the power by 7.7 percent from the design of the 'Valid Prototype', System80+. Thus, reliable evaluations of potential adverse flow effects on the steam generator of APR1400 are necessary according to the regulatory guide. This paper is part of the computational fluid dynamics (CFD) analysis results for evaluation of the adverse flow effect for the steam generator internals of APR1400, including a series of sensitivity analyses to enhance the reliability of CFD analysis and an estimation the effect of flow loads on the internals of the steam generator under normal operation conditions

Allothermal steam gasification of biomass in cyclic multi-compartment bubbling fluidized-bed gasifier/combustor - new reactor concept.

Science.gov (United States)

Iliuta, Ion; Leclerc, Arnaud; Larachi, Faïçal

2010-05-01

A new reactor concept of allothermal cyclic multi-compartment fluidized bed steam biomass gasification is proposed and analyzed numerically. The concept combines space and time delocalization to approach an ideal allothermal gasifier. Thermochemical conversion of biomass in periodic time and space sequences of steam biomass gasification and char/biomass combustion is simulated in which the exothermic combustion compartments provide heat into an array of interspersed endothermic steam gasification compartments. This should enhance unit heat integration and thermal efficiency and procure N(2)-free biosyngas with recourse neither to oxygen addition in steam gasification nor contact between flue and syngas. The dynamic, one-dimensional, multi-component, non-isothermal model developed for this concept accounts for detailed solid and gas flow dynamics whereupon gasification/combustion reaction kinetics, thermal effects and freeboard-zone reactions were tied. Simulations suggest that allothermal operation could be achieved with switch periods in the range of a minute supporting practical feasibility for portable small-scale gasification units. Copyright 2009 Elsevier Ltd. All rights reserved.

Steam and sodium leak simulation in a fluidized-bed steam generator

International Nuclear Information System (INIS)

Vaux, W.G.; Keeton, A.R.; Keairns, D.L.

1977-01-01

A fluidized-bed steam generator for the liquid metal fast breeder reactor enhances plant availability and minimizes the probability of a water/sodium reaction. An experimental test program was conceived to assess design criteria and fluidized-bed operation under conditions of water, steam, and sodium leaks. Sodium, steam, and water were leaked into helium-fluidized beds of metal and ceramic particles at 900 F. Test results show the effects of leaks on the heat transfer coefficient, quality of fluidization, leak detection, and cleanup procedures

SCW Pressure-Channel Nuclear Reactor Some Design Features

Science.gov (United States)

Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

Experimental and analytical investigations to air and steam ingress into the vacuum vessel of fusion reactors

International Nuclear Information System (INIS)

Kruessenberg, A.K.

1996-12-01

The basic fusion safety objective is the development of fusion power plants with features that protect individuals, society and the environment by establishing and maintaining an effective defence against radiological and other hazards. The most important specific principle is the establishment of three sequential levels of defence, characterized in priority order by prevention, protection and mitigation. The safety conscious selection of materials as one prevention feature gives the basis for the work described in this report. In order to protect the metallic first wall of fusion reactors from direct interaction with the plasma an extra armour is foreseen. Carbon offers the features low atomic number, high melting point, high thermal conductivity and good mechanical stability up to high temperatures making it to a favourite armour material. Looking on the safety behaviour of fusion reactors it has to be noted that carbon is unstable against oxidizing media like oxygen and steam at high temperatures und carbon has a high sorption capacity for radiologically important tritium. And tritium used as intermediate fuel in the actual reactor concepts is the one form radioactivity is present in fusion reactors. Accidents like loss of vacuum (LOVA) will lead to an air ingress into the vacuum vessel, oxidation of the hot carbon and a partial mobilization of the sorbed tritium. In a similar manner loss of coolant into vacuum (LOCIV) will lead to a water/steam ingress into the vacuum vessel, also accompanied by carbon oxidation and tritium release. (orig.)

Main Steam Line Break Analysis for the Fully Passive Safety System of SMART

International Nuclear Information System (INIS)

Kim, Seong Wook; Chun, Ji Han; Bae, Kyoo Hwan; Kim, Keung Koo

2013-01-01

The standard design approval of SMART (System-integrated Modular Advanced ReacTor) developed by KAERI and KEPCO consortium was issued on July 4, 2012. Although SMART has enhanced safety compared to the conventional reactor, there is a demand to meet the 'passive safety performance requirements' after the Fukushima accident. The passive safety performance requirements are the capabilities to maintain the plant at a safe shutdown condition for a minimum of 72 hours without AC power supply or operator action in case of design basis accident (DBA). To satisfy the requirements, KAERI is developing a safety enhanced SMART by adopting a passive safety injection system. The passive safety injection system developed for SMART is a gravity-driven injection system, which consists of four trains, each of which includes a pressure balance line, core makeup tank (CMT), safety injection tank (SIT) and injection line. The CMT plays an important role to inject borated water into the RCS to prevent or dissolve the return to power (re-criticality) condition during the event of increase in heat removal by the secondary system. The main steam line break accident (MSLB) is the most limiting accident for an increase in heat removal by the secondary system. In this study, the safety analysis results of MSLBs at hot full power condition and at hot zero power condition in view of re-criticality are given. The MSLB accident has been analyzed for the SMART adopting fully passive safety system in the aspect of re-criticality. The results show that the core remains subcritical condition throughout the transient due to the borated water injected by the CMT. As further works, many kinds of analyses and sensitivity studies should be performed for the design establishment and improvement of the fully passive system of SMART

Ethanol steam reforming kinetics of a Pd-Ag membrane reactor

Energy Technology Data Exchange (ETDEWEB)

Tosti, Silvano; Borelli, Rodolfo; Borgognoni, Fabio [ENEA, Dipartimento FPN, C.R. ENEA Frascati, Via E. Fermi 45, Frascati (RM) I-00044 (Italy); Basile, Angelo [Institute on Membrane Technology, ITM-CNR, c/o Univ. of Calabria, via P. Bucci, Cubo 17/C, 87030 Rende (CS) (Italy); Castelli, Stefano [ENEA, Dipartimento ACS, C.R. ENEA Casaccia, Via Anguillarese 301, Roma I-00123 (Italy); Fabbricino, Massimiliano; Licusati, Celeste [Dept. of Hydraulic and Environmental Engineering, Univ. of Naples Federico II, Via Claudio 21, Naples 80125 (Italy); Gallucci, Fausto [Fundamentals of Chemical Reaction Engineering Group, Faculty of Science and Technology, University of Twente, Enschede (Netherlands)

2009-06-15

The ethanol steam reforming reaction carried out in a Pd-based tubular membrane reactor has been modelled via a finite element code. The model considers the membrane tube divided into finite volume elements where the mass balances for both lumen and shell sides are carried out accordingly to the reaction and permeation kinetics. Especially, a simplified ''power law'' has been applied for the reaction kinetics: the comparison with experimental data obtained by using three different kinds of catalyst (Ru, Pt and Ni based) permitted defining the coefficients of the kinetics expression as well as to validate the model. Based on the Damkohler-Peclet analysis, the optimization of the membrane reformer has been also approached. (author)

Some engineering aspects of the steam generator system for the United States LMFBR demonstration plant

International Nuclear Information System (INIS)

Tippets, F.E.

1975-01-01

This paper describes the main design features of the steam generator system for the Clinch River Breeder Reactor Plant and the engineering approach being employed for some of the critical elements of this system, including in particular the sodium-steam/water boundary, the efforts to have this boundary be of highest integrity, and the system features to safely accommodate any failure of the boundary. (author)

Some engineering aspects of the steam generator system for the United States LMFBR demonstration plant

Energy Technology Data Exchange (ETDEWEB)

Tippets, F E

1975-07-01

This paper describes the main design features of the steam generator system for the Clinch River Breeder Reactor Plant and the engineering approach being employed for some of the critical elements of this system, including in particular the sodium-steam/water boundary, the efforts to have this boundary be of highest integrity, and the system features to safely accommodate any failure of the boundary. (author)

Heat extraction from HTGR reactor

International Nuclear Information System (INIS)

Balajka, J.; Princova, H.

1986-01-01

The analysis of an HTGR reactor energy balance showed that steam reforming of natural gas or methane is the most suitable process of utilizing the high-temperature heat. Basic mathematical relations are derived allowing to perform a general energy balance of the link between steam reforming and reactor heat output. The results of the calculation show that the efficiency of the entire reactor system increases with increasing proportion of heat output for steam reforming as against heat output for the steam generator. This proportion, however, is limited with the output helium temperature from steam reforming. It is thus always necessary to use part of the reactor heat output for the steam cycle involving electric power generation or low-potential heat generation. (Z.M.)

Hydrodynamic and acoustic analysis in 3-D of a section of main steam line to EPU conditions; Analisis hidrodinamico y acustico en 3D de una seccion de linea de vapor principal a condiciones de EPU

Energy Technology Data Exchange (ETDEWEB)

Centeno P, J.; Castillo J, V.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico); Nunez C, A.; Polo L, M. A., E-mail: baldepeor21@hotmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2013-10-15

The objective of this word is to study the hydrodynamic and acoustic phenomenon in the main steam lines (MSLs). For this study was considered the specific case of a pipe section of the MSL, where is located the standpipe of the pressure and/or safety relief valve (SRV). In the SRV cavities originates a phenomenon known as whistling that generates a hydrodynamic disturbance of acoustic pressure waves with different tones depending of the reactor operation conditions. In the SRV cavities the propagation velocity of the wave can originate mechanical damage in the structure of the steam dryer and on free parts. The importance of studying this phenomenon resides in the safety of the integrity of the reactor pressure vessel. To dissipate the energy of the pressure wave, acoustic side branches (ASBs) are used on the standpipe of the SRVs. The ASBs are arrangements of compacted lattices similar to a porous medium, where the energy of the whistling phenomenon is dissipate and therefore the acoustic pressure load that impacts in particular to the steam dryers, and in general to the interns of the vessel, diminishes. For the analysis of the whistling phenomenon two three-dimensional (3-D) models were built, one hydrodynamic in stationary state and other acoustic for the harmonic times in transitory regimen, in which were applied techniques of Computational Fluid Dynamics. The study includes the reactor operation analysis under conditions of extended power up rate (EPU) with ASB and without ASB. The obtained results of the gauges simulated in the MSL without ASB and with ASB, show that tones with values of acoustic pressure are presented in frequency ranges between 160 and 200 Hz around 12 MPa and of 7 MPa, respectively. This attenuation of tones implies the decrease of the acoustic loads in the steam dryer and in the interns of the vessel that are designed to support pressures not more to 7.5 MPa approximately. With the above-mentioned is possible to protect the steam dryer

Dynamic load in suppression pool during BWR main steam safety relief valve actuation

International Nuclear Information System (INIS)

Tsukada, Hiroshi; Yamaguchi, Hirokatsu; Morita, Terumichi

1979-01-01

BWRs are so designed that the exhaust steam from main steam safety relief valves is led to pressure suppression pools, and the steam is condensed in pool water, but at this time, dynamic load seems to arise in the pool water. In Tokai No. 2 Power Station, a Mark-2 containment vessel was adopted to improve the reliability as much as possible and to obtain the design with margin. In this report, the result of actual machine test in Tokai No. 2 Power Station and the method of reducing the load are described. When a relief valve works, the discharge of water in exhaust pipes into a suppression pool, the exhaust of air in exhaust pipes and repeated expansion and contraction of bubbles in pool water, and the exhaust of steam and condensation occur. As for the construction of the suppression pool in Tokai No. 2 Power Station, cross-shaped quencher and the structure with jet deflector were installed. The test plan and the test result with an actual machine are reported. The soundness of the Mark-2 containment vessel and the structures in the pool was proved. The differential pressure acting on the structures was negligibly small. The measured pulsating pressure was in the range from 0.84 to -0.39 kg/cm 2 . (Kako, I.)

Opinion on serviceability of Bugey 3 reactor steam generators until their replacement foreseen in September 2010

International Nuclear Information System (INIS)

2010-04-01

This document briefly reports the damage characterization of tubular bundles in steam generators of the Bugey 3 reactor, discusses the actions which are foreseen to prevent a tube failure risk, and discusses the risk of leakage during operation. Recommendations are formulated about investigation on the corrosion, and about prediction computation to be performed

A Study on the Main Steam Safety Valve Opening Mechanism by Flashing on NPPs

International Nuclear Information System (INIS)

Kim, Bae Joo

2009-01-01

A safety injection event happened by opening of the Main Steam Safety Valve at Kori unit 1 on April 16, 2005. The safety valves were opened at the lower system pressure than the valve opening set point due to rapid system pressure drop by opening of the Power Operated Relief Valve installed at the upstream of the Main Steam System. But the opening mechanism of safety valve at the lower set point pressure was not explained exactly. So, it needs to be understood about the safety valve opening mechanism to prevent a recurrence of this kind of event at a similar system of Nuclear Power Plant. This study is aimed to suggest the hydrodynamic mechanism for the safety valve opening at the lower set point pressure and the possibility of the recurrence at similar system conditions through document reviewing for the related previous studies and Kori unit 1 event

Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

International Nuclear Information System (INIS)

Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R.

2008-03-01

This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to the

Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

Energy Technology Data Exchange (ETDEWEB)

Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R. [Royal Institute of Technology (KTH), (Sweden)

2008-03-15

This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to

Thermal-hydraulics of wave propagation and pressure distribution under hypothetical steam explosion conditions in the ANS reactor

Energy Technology Data Exchange (ETDEWEB)

Taleyarkhan, R.P.; Georgevich, V.; N-Valenit, S.; Kim, S.H. [Oak Ridge National Lab., TN (United States)

1995-09-01

This paper describes salient aspects of the modeling and analysis framework for evaluation of dynamic loads, wave propagation, and pressure distributions (under hypothetical steam explosion conditions) around key structural boundaries of the Advanced Neutron Source (ANS) reactor core region. A staged approach was followed, using simple thermodynamic models for bounding loads and the CTH code for evaluating realistic estimates in a staged multidimensional framework. Effects of nodalization, melt dispersal into coolant during explosion, single versus multidirectional dissipation, energy level of melt, and rate of energy deposition into coolant were studied. The importance of capturing multidimensional effects that simultaneously account for fluid-structural interactions was demonstrated. As opposed to using bounding loads from thermodynamic evaluations, it was revealed that the ANS reactor system will not be vulnerable to vertically generated missiles that threaten containment if realistic estimates of energetics are used (from CTH calculations for thermally generated steam explosions without significant aluminum ignition).

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Main results of assessing integrity of RNPP-3 steam generator heat exchange tubes in accident management

International Nuclear Information System (INIS)

Shugajlo, Al-j P.; Mustafin, M.A.; Shugajlo, Al-r P.; Ryzhov, D.I.; Zhabin, O.I.

2017-01-01

Tubes integrity evaluation under accident conditions considering drain of SG and current technical state of steam exchange tubes is an important question regarding SG long-term operation and improvement of accident management strategy.The main investigation results prepared for heat exchange surface of RNPP-3 steam generator are presented in this research aimed at assessing integrity of heat exchange tubes under accident conditions, which lead to full or partial drain of heat exchange surface, in particular during station blackout.

Steam generator tube performance: world experience with water-cooled nuclear power reactors during 1979

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1981-01-01

The performance of steam generator tubes in water-cooled nuclear power reactors is reviewed for 1979. Tube failures occurred at 38 of the 93 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The defect rate, although higher than that in 1978, was still lower than the rates of the two previous years. Methods being employed to detect defects include the increased use of multifrequency eddy-current testing and a trend to full-length inspection of all tubes. To reduce the incidence of tube failure by corrosion, plant operators are turning to full-flow condensate demineralization and more leak-resistant condenser tubes. 10 tables

IAEA activities on steam generator life management

International Nuclear Information System (INIS)

Gueorguiev, B.; Lyssakov, V.; Trampus, P.

2002-01-01

The International Atomic Energy Agency (IAEA) carries out a set of activities in the field of Nuclear Power Plant (NPP) life management. Main activities within this area are implemented through the Technical Working Group on Life Management of NPPs, and mostly concentrated on studies of understanding mechanisms of degradation and their monitoring, optimisation of maintenance management, economic aspects, proven practices of and approaches to plant life management including decommissioning. The paper covers two ongoing activities related to steam generator life management: the International Database on NPP Steam Generators and the Co-ordinated Research Project on Verification of WWER Steam Generator Tube Integrity (WWER is the Russian designed PWR). The lifetime assessment of main components relies on an ability to assess their condition and predict future degradation trends, which to a large extent is dependent on the availability of relevant data. Effective management of ageing and degradation processes requires a large amount of data. Several years ago the IAEA started to work on the International Database on NPP Life Management. This is a multi-module database consisting of modules such as reactor pressure vessels materials, piping, steam generators, and concrete structures. At present the work on pressure vessel materials, on piping as well as on steam generator is completed. The paper will present the concept and structure of the steam generator module of the database. In countries operating WWER NPPs, there are big differences in the eddy current inspection strategy and practice as well as in the approach to steam generator heat exchanger tube structural integrity assessment. Responding to the need for a co-ordinated research to compare eddy current testing results with destructive testing using pulled out tubes from WWER steam generators, the IAEA launched this project. The main objectives of the project are to summarise the operating experiences of WWER

Conceptual design study of Cu bonded steam generator

International Nuclear Information System (INIS)

Chikazawa, Yoshitaka; Konomura, Mamoru

2004-05-01

In phase II of feasibility study of commercialized fast reactor cycle systems of JNC, we make a concept of a sodium cooled reactor without secondary sodium circuits. And a sodium cooled reactor with Cu bonded steam generators is one of promising concept. As the result of FY 2001 study, the construction cost of reactor cooling system with rectangular tube Cu bonded steam generators is 0.71 to 1.23 times as much as that of an ordinary sodium cooled reactor with secondary sodium circuits. In the FY 2003 study, plastic and creep analysis to evaluate life distortion are carried out and inelastic strains and creep fatigue damage are checked for full code compliance. The NNC's crack growth experiments show that there are few possibility to penetrate a crack from the steam tube side to the sodium tube side at the operating temperature. But penetration is observed in a four point bend test at the room temperature, because the notch opens widely in the bend test. In the FY 2004 study, a gas pressurized crack growth experiment is planed to confirm that there is no crack penetration in the condition of operating steam generators. (author)

Vibrations as a contributor to the cracking of PGV-1000 steam generator cold collector

International Nuclear Information System (INIS)

Verezemskij, V.G.

1997-01-01

The results of multiple investigations of cold collector ruptures at NPPs with WWER-1000 reactor as a complex and multi-parameter process are generalized. It is shown that the temperature of cold collector operation (280-290 deg C) at which environment corrosion effects are maximum has an important role for revealing the factors causing the damage. When the reactor plant operates under stationary and transient conditions the primary coolant circuit equipment, main circulation pipelines and main steam lines become involved into vibrations with different intensities as a result of pressure pulsations at reverse and multiple frequencies of the main circulation pumps connected with formation of standing pressure waves in the primary circuit and steam lines. The analysis made gives an opportunity to conclude that dynamic loads (vibrations) play the leading part in reaching the limits of cold collector metal cyclic strength and its cracking. It means that the measures for cold collector service life increasing should be directed on vibration amplitude lowering and cyclic stress decreasing

Determination of stresses caused by fluctuation of acoustic load in the steam dryers of a BWR

International Nuclear Information System (INIS)

Centeno P, J.; Quezada G, S.; Prieto G, A.; Vazquez R, A.; Espinosa P, G.; Nunez C, A.

2014-10-01

The extended power up-rate (EPU) in a nuclear power plant cause various problems in BWR components also in the steam system. This due to increased steam flow generated in the reactor and is conveyed to the turbine by the four main steam lines (MSL). One of the most serious problems is the generation of acoustic pressure loads in the metal structure of the steam dryer which eventually leads to fatigue failure and even the appearance of cracks, and in turn it causes loose parts that are entrained by the steam and transported in the MSL. This problem is due to the fluctuation of load acoustics caused by the union of the safety or relief valves (SRV) with the MSL, spreading through these to reach the reactor pressure vessel (RPV) where the effect of resonance of the acoustic wave is amplified and impacts directly in the supporting structure of the steam dryer, skirt and the panels where the mixture liquid-steam is dried, by centrifugation effect and runoff of liquid water. Efforts in the steam dryer operating conditions of EPU for two cases will be analyzed in this work, the first is before the installation of Acoustic Side Branch (ASB), and in the second case we consider the installation of said ASB in the standpipes of SRV. The analysis was performed with numerical experiments on a platform for computational fluid dynamics with virtual geometries previously designed based on the actual components of the reactor and steam system. The model to study is delimited by the top of the RPV, the steam dryer and a section of each of the four MSL with ten standpipes of SRV. With the obtained data and considering the mechanical-structural properties of the steam dryer material, we can evaluate the mechanical resistance to impacts by acoustic pressure load and its possible deformation or cracking. (Author)

High-temperature reactors. Activities in France on the steam cycle HTR

International Nuclear Information System (INIS)

Lacoste Lareymondie, de; Guennec, N.; Rastoin, J.

1975-01-01

Although French activities cover all the possibilities of high-temperature reactors the effort of the last few years has been concentrated on the steam cycle electricity-generating version. This work, closely coordinated with that of General Atomic in application of agreements settled in 1972 and 1973, was devoted to engineering as a result of the assimilation of American technique by French industry and to research and development owing to the joint CEA and GA programme. After an examination of these two centers of activity the reasons which will lead to a closer collaboratin among the European partners of General Atomic are expressed in conclusion [fr

Maintenance and repair of LMFBR steam generators: specialists` meeting, O-Arai Engineering Center, Japan, 4-8 June 1984. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1984-07-01

The Specialists` Meeting on "Maintenance and Repair of LMFBR Steam Generators" was held in Oarai, Japan, from 4-8 June 1984. The meeting was sponsored by the International Atomic Energy Agency on the recommendation of the IAEA International Working Group on Fast Reactors and was hosted by the Power Reactor and Nuclear Fuel Development Corporation of Japan. The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topical areas were discussed by participants: national review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; research and development work on maintenance and repair; and experience on steam generator maintenance and repair.

Main results of BN-600 reactor stress-strain state investigations

International Nuclear Information System (INIS)

Panov, V.A.

1983-01-01

The development of BN-600 fast reactor plant needed the solution of a series of complex engineering problems including ones for confirming integrity of the most vital structural components. The particular attention was given to the main vessel since reactor availability end safe operation of the plant as a whole depend on vessel strength end integrity. The present report deals with the main results of theoretical and experimental investigations of the stress-strain state of BN-600 reactor vessel carried out during design, start-up and initial bringing the reactor to power

Reactor container cooling device

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1995-11-10

The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

Tachometric flowmeters for measuring circulation water parameters in steam generators of the NPPs running on pressurized water reactors

International Nuclear Information System (INIS)

Ageev, A.G.; Korolkov, B.M.; Nigmatulin, B.I.; Belov, V.I.; Vasileva, R.V.; Trubkin, N.I.

1997-01-01

Tachometric flowmeters used in steam generators for determining the velocity and direction of the flow have a limited service life owing to the use of corundum for the bearing assembly components. Various materials were investigated for the feasibility of using them as alternatives for replacing the corundum bearing and guide bushing under conditions close to the conditions in steam generators: 7 MPa, 260 degC. Good results were obtained with bearing assemblies fabricated from corrosion-resistant steel. Testing of the transducer design and optimization of the technique was accomplished in the course of testing steam generators of the WWER-1000 reactor at the Balakovskaya nuclear power plant. The velocity and direction of flow in the steam generator were measured within a wide range of unit power ratings up to the values corresponding to full power output. The service life of the transducers proved to be not less than 720 hours. The transducer parameters remained unchanged over the entire operation period. (M.D.)

R.B. pressure and temperature transient following main steam line break

International Nuclear Information System (INIS)

Das, M.; Bhawal, R.N.; Prakash, P.

1989-01-01

The R.B. containment plays an important role in mitigating the consequences of any accident core. The analysis of Main Steam Line Break (MSLB), though not of relevance from activity release considerations, is essentially from structural integrity point of view. In this paper the outline of the likely scenario is drawn and the approach for thermal hydraulic simulation of the system for carrying out transient blowdown analysis is discussed. The results of the containment pressure and temperature transient analysis are also presented. (author). 4 refs., 7 figs

Occupational analysis for the Angra-1 reactor

International Nuclear Information System (INIS)

Moraes, A.

1991-01-01

Due to several modifications which were imposed to its time schedule during construction, the Angra-1 reactor did not enter to the grid in 1982 as it was initially foreseen. These modifications occurred due to an unforeseen scenario that was verified in steam generators (serie D-3, Westinghouse) of power stations with similar configurations which had been installed in other countries such as Ringhals-3 (Sweden), Almaraz-1 (Spain) and McGuine-1 (USA). So, among the main events that occurred in the Angra-1 reactor, which were of interest from the point of view of radiation protection, it could be pointed out the personnel monitoring, and the occupational exposure measurements at different reactor power, during the reactor fueling and during modification and tests performed at the steam generators and at ducts of the primary coolant circuit. (author)

Reconstruction of steam generators super emergency feadwater supply system (SHNC) and steam dump stations to the atmosphere system PSA

International Nuclear Information System (INIS)

Kuzma, J.

2001-01-01

Steam Generators Super Emergency Feadwater Supply System (SHNC) and Steam Dump Stations to the Atmosphere System (PSA) are two systems which cooperate to remove residual heat from reactor core after seismic event. SHNC assure feeding of the secondary site of steam generator (Feed) where after heat removal.from primary loops, is relieved to the atmosphere by PSA (Bleed) in form of steam. (author)

Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

International Nuclear Information System (INIS)

Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo

2016-01-01

Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate

Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

Energy Technology Data Exchange (ETDEWEB)

Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate.

Regulatory analysis for the resolution of generic issue C---8, main steam isolation valve leakage and LCS [leakage control system] failure

International Nuclear Information System (INIS)

Graves, C.C.

1990-06-01

Generic Issue C-8 deals with staff concerns about public risk because of the incidence of leak test failures reported for main steam isolation valves (MSIVs) at boiling water reactors and the limitations of the leakage control systems (LCSs) for mitigating the consequences of leakage from these valves. If the MSIV leakage is greatly in excess of the allowable value in the technical specifications, the LCS would be unavailable because of design limitations. The issue was initiated in 1983 to assess (1) the causes of MSIV leakage failures, (2) the effectiveness of the LCS and alternative mitigation paths, and (3) the need for additional regulatory action to reduce public risk. This report presents the regulatory analysis for Generic Issue C-8 and concludes that no new regulatory requirements are warranted

Numerical Analysis on Transient of Steam-gas Pressurizer

International Nuclear Information System (INIS)

Kim, Jong-Won; Lee, Yeon-Gun; Park, Goon-Cherl

2008-01-01

In nuclear reactors, various pressurizers are adopted to satisfy their characteristics and uses. The additional active systems such as heater, pressurizer cooler, spray and insulator are essential for a steam or a gas pressurizer. With a steam-gas pressurizer, additional systems are not required due to the use of steam and non-condensable gas as pressure-buffering materials. The steam-gas pressurizer in integrated small reactors experiences very complicated thermal-hydraulic phenomena. To ensure the integrity of this pressurizer type, the analysis on the transient behavior of the steam-gas pressure is indispensable. For this purpose, the steam-gas pressurizer model is introduced to predict the accurate system pressure. The proposed model includes bulk flashing, rainout, inter-region heat and mass transfer and wall condensation with non-condensable gas. However, the ideal gas law is not applied because of significant interaction at high pressure between steam and non-condensable gas. The results obtained from this proposed model agree with those from pressurizer tests. (authors)

Advanced designs of VVER reactor plant

International Nuclear Information System (INIS)

Mokhov, V.A.

2010-01-01

The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

Reactor container

International Nuclear Information System (INIS)

Furukawa, Hideyasu; Oyamada, Osamu; Uozumi, Hiroto.

1976-01-01

Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)

Steam separator-superheater with drawing of a fraction of the dried steam

International Nuclear Information System (INIS)

Bessouat, Roger; Marjollet, Jacques.

1976-01-01

This invention concerns a vertical separator-superheater of the steam from a high pressure expansion turbine before it is admitted to an expansion turbine at a lower pressure, by heat exchange with steam under a greater pressure, and drawing of a fraction of the dried steam before it is superheated. Such drawing off is necessary in the heat exchange systems of light water nuclear reactors. Its purpose is to provide a separator-superheater that provides an even flow of non superheated steam and a regular distribution of the steam to be superheated to the various superheating bundles, with a significantly uniform temperature of the casing, thereby preventing thermal stresses and ensuring a minimal pressure drop. The vertical separator-superheater of the invention is divided into several vertical sections comprising as from the central area, a separation area of the steam entrained water and a superheater area and at least one other vertical section with only a separation area of the steam entrained water [fr

BWR type reactor system

International Nuclear Information System (INIS)

Morooka, Shin-ichi.

1980-01-01

Purpose: To reduce the internal structure in a reactor by rapidly and efficiently transferring heat generated in a reactor core out of the reactor and eliminating the danger of radiation exposure. Constitution: Steam generated in a pressure vessel is introduced into heat pipe group by inserting the heat pipe group into the steam dome of the pressure vessel. The introduced steam is condensed in the heat pipes to transfer the heat of the steam to the heat pipe group. The transferred heat is transmitted to a heat exchanger provided out of a containment vessel to generate steam to operate a turbine. Thus, it is not necessary to introduce the steam including radioactive substance externally and can remove only the heat so as to carry out the desired purpose. (Kamimura, M.)

Burst protected nuclear reactor plant with PWR

International Nuclear Information System (INIS)

Harand, E.; Michel, E.

1978-01-01

In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

Issues in the selection of the LMFBR steam cycle

International Nuclear Information System (INIS)

Buschman, H.W.; McConnell, R.J.

1983-01-01

Unlike the light-water reactor, the liquid-metal fast breeder reactor (LMFBR) allows the designer considerable latitude in the selection of the steam cycle. This latitude in selection has been exercised by both foreign and domestic designers, and thus, despite the fact that over 25 LMFBR's have been built or are under construction, a consensus steam cycle has not yet evolved. This paper discusses the LMFBR steam cycles of interest to the LMFBR designer, reviews which of these cycles have been employed to date, discusses steam-cycle selection factors, discusses why a consensus has not evolved, and finally, concludes that the LMFBR steam-cycle selection is primarily one of technical philosophy with several options available

Development of main steam safety valve set pressure evaluating system

International Nuclear Information System (INIS)

Oketani, Koichiro; Manabe, Yoshihisa.

1991-01-01

A main steam safety valve set pressure test is conducted for all valves during every refueling outage in Japan's PWRs. Almost all operations of the test are manually conducted by a skilled worker. In order to obtain further reliability and reduce the test time, an automatic test system using a personnel computer has been developed in accordance with system concept. Quality assurance was investigated to fix system specifications. The prototype of the system was manufactured to confirm the system reliability. The results revealed that this system had high accuracy measurement and no adverse influence on the safety valve. This system was concluded to be applicable for actual use. (author)

Structural analysis of steam generator internals following feed water main steam line break: DLF approach

International Nuclear Information System (INIS)

Bhasin, Vivek; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

1993-01-01

In order to evaluate the possible release of radioactivity in extreme events, some postulated accidents are analysed and studied during the design stage of Steam Generator (SG). Among the various accidents postulated, the most important are Feed Water Line Break (FWLB) and Main Steam Line Break (MSLB). This report concerns with dynamic structural analysis of SG internals following FWLB/MSLB. The pressure/drag-force time histories considered were corresponding to the conditions leading to the accident of maximum potential. The SG internals were analysed using two approaches of structural dynamics. In first approach simplified DLF method was adopted. This method yields an upper bound values of stresses and deflection. In the second approach time history analysis by Mode Superposition Technique was adopted. This approach gives more realistic results. The structure was qualified as per ASME B and PV Code SecIII NB. It was concluded that in all the components except perforated flow distribution plate, the stress values based on elastic analysis are within the limits specified by ASME Code. In case of perforated flow distribution plate during the MSLB transient the stress values based on elastic analysis are higher than the ASME Code limits. Therefore, its limit load analysis had to be done. Finally, the collapse pressure evaluated using limit load analysis was shown to be within the limits of ASME B and PV Code SecIII Nb. (author). 31 refs., 94 figs., 16 tabs

Possibilities of the metallurgical base in the manufacture of tubes for nuclear power plant steam generators

International Nuclear Information System (INIS)

Prnka, T.; Walder, V.; Dolenek, J.

Current possibilities are briefly summarized of metallurgy in the manufacture of high-quality tubes for nuclear power plant steam generators, mainly for fast reactor power plants. Discussed are steel making possibilities, semi-finished product and tube forming with special regard to 2.25Cr1MoNiNb steel problems, heat treatment, finishing, and testing. Necessary equipment and technology for the production of steam generator tubes are less common in the existing practice and are demanding on investment; their introduction, however, is inevitable for securing quality production of steam generator tubes. (Kr)

Water chemistry in boiling water reactors - A Leibstadt-specific overview

International Nuclear Information System (INIS)

Sarott, F.-A.

2005-01-01

The boiling water reactor (BWR) consists of two main water circuits: the water-steam cycle and the main cooling water system. In the introduction, the goals and tasks of the BWR plant chemistry are described. The most important objectives are the prevention of system degradation by corrosion and the minimisation of radiation fields. Then a short description of the BWR operation principle, including the water steam cycle, the transport of various impurities by the steam, removing impurities from the condensate, the reactor water clean-up system, the balance of plant and the main cooling water system, is given. Subsequently, the focus is set on the water-steam cycle chemistry. In order to fulfil the somewhat contradictory requirements, the chemical parameters must be well balanced. This is achieved by the water chemistry control method called 'normal water chemistry'. Other additional methods are used for the solution to different problems. The 'zinc addition method' is applied to reduce high radiation levels around the recirculation loops. The 'hydrogen water chemistry method' and the 'noble metal chemical addition method' are used to protect the reactor core components and piping made of stainless steel against stress corrosion cracking. This phenomenon has been observed for about 40 years and is partly due to the strong oxidising conditions in the BWR water. Both mitigation methods are used by the majority of the BWR plants all over the world (including the two Swiss NPPs Muehleberg and Leibstadt). (author)

Steam generator tube performance: experience with water-cooled nuclear power reactors during 1983 and 1984

International Nuclear Information System (INIS)

Tatone, O.S.; Meindl, P.; Taylor, G.F.

1986-06-01

A review of the performance of steam generator tubes in water-cooled nuclear power reactors showed that tubes were plugged at 47 (35.6%) of the reactors in 1983 and at 63 (42.6%) of the reactors during 1984. In 1983 and 1984 3291 and 3335 tubes, respectively, were removed from service, about the same as in 1982. The leading causes assigned to tube failure were stress corrosion cracking from the primary side and stress corrosion cracking or intergranular attack from the secondary side. In addition 5668 tubes were repaired for further service by installation of internal sleeves. Most of these were believed to have deteriorated by one of the above mechanisms or by pitting. There is a continuing trend towards high-integrity condenser tube materials at sites cooled by brackish or sea water. 31 refs

Improvement of steam separator in boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Schaefer, Jan Peter; Cremer, Ingo; Lorenz, Maik [AREVA GmbH, Erlangen (Germany)

2013-07-01

The potential to improve the function of the steam separator is identified and explored by scaled air-water tests and validated CFD calculations. It can be outlined that requirements on a modern steam separator for BWR plants will be fulfilled, combined with very good operational experience of the existing separator designs (e.g. material, layout). With the new steam separator design, modern high performance fuel assembly designs can be used for various core loading strategies (e.g. low leakage). This allows an increased thermal power of up to +50 % for the fuel element clusters in the center of the core with high radial peaking factors. In addition, any problems with unallowable high moisture at the turbine are solved with the new design, which have been identified for running BWR plants with the old steam separator design after applying new core loading patterns (e.g. after power uprates). A compatible steam separator design for all running BWRs is ready to launch. (orig.)

Coupling of high temperature nuclear reactor with chemical plant by means of steam loop with heat pump

Directory of Open Access Journals (Sweden)

Kopeć Mariusz

2017-01-01

Full Text Available High temperature nuclear reactors (HTR can be used as an excellent, emission-free source of technological heat for various industrial applications. Their outlet helium temperature (700°-900°C allows not only for heat supply to all processes below 600°C (referred to as “steam class”, but also enables development of clean nuclear-assisted hydrogen production or coal liquefaction technologies with required temperatures up to 900°C (referred to as “chemical class”. This paper presents the results of analyses done for various configurations of the steam transport loop coupled with the high-temperature heat pump designed for “chemical class” applications. The advantages and disadvantages as well as the key issues are discussed in comparison with alternative solutions, trying to answer the question whether the system with the steam loop and the hightemperature heat pump is viable and economically justified.

Thermohydraulic feedbacks in self-pressurized reactor systems

International Nuclear Information System (INIS)

Fiebig, R.

1977-01-01

The impact on the dynamic behaviour of a self-pressurized reactor by the thermodynamic properties of the steam dome is investigated. For self-stabilization of the system the water of the primary circuit must be coupled thermodynamically to the steam in the steam dome, or alternatively the water in the reactor core must be subcooled sufficiently. Ways of thermodynamically coupling the water to the steam are heat conduction, boiling and condensation. A heat sink within the steam dome forces thermodynamic equilibrium between water and steam. This condition yields excellent self-control. Without heat sink thermal coupling is suspended at transients resulting in pressure rises. However, the reactor is still controlable as long as circuit and steam dome have direct contact. At the reactor of the NCS-80 a buffer volume of water separates primary circuit and steam volume. Stability is achieved by a heat sink in the steam dome and a shift of the core temperature into the subcooled domain effected by steam bubbles rising into the steam dome. (orig.) [de

Thermohydraulic feedbacks in self-pressurized reactor systems

International Nuclear Information System (INIS)

Fiebig, R.

1977-01-01

The impact on the dynamic behaviour of a self-pressurized reactor by the thermodynamic properties of the steam dome is investigated. For self-stabilization of the system the water of the primary circuit must be coupled thermodynamically to the steam in the steam dome, or alternatively the water in the reactor core must be subcooled sufficiently. Ways of thermodynamically coupling the water to the steam are heat conduction, boiling and condensation. A heat sink within the steam dome forces thermodynamic equilibrium between water and steam. This condition yields excellent self-control. Without heat sink thermal coupling is suspended at transients resulting in pressure rises. However, the reactor is still controllable as long as circuit and steam dome have direct contact. At the reactor of the NCS-80 a buffer volume of water separates primary circuit and steam volume. Stability is achieved by a heat sink in the steam dome and a shift of the core temperature into the subcooled domain effected by steam bubbles rising into the steam dome. (orig.) [de

Considerations in selecting tubing materials for CANDU steam generators

International Nuclear Information System (INIS)

Hemmings, R.L.

1978-01-01

Corrosion resistance is the major consideration in selecting tubing material for CANDU steam generators. Corrosion, and additional considerations, lead to the following steam generator tubing material recommendations: for CANDU-BPHWR's (boiling pressurized heavy water reactors) low-cobalt Incoloy-800; for CANDU-PHWR's (pressurized, non-boiling, heavy water reactors), low-cobalt Monel-400

Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

Energy Technology Data Exchange (ETDEWEB)

Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Prieto-Guerrero, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Núñez-Carrera, A. [Comisión Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragán 779, Col. Narvarte, México, D.F. 03020 (Mexico); Vázquez-Rodríguez, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Centeno-Pérez, J. [Instituto Politécnico Nacional, Escuela Superior de Física y Matemáticas Unidad Profesional “Adolfo López Mateos”, Av. IPN, s/n, México, D.F. 07738 (Mexico); Espinosa-Martínez, E.-G. [Departamento de Sistemas Energéticos, Universidad Nacional Autónoma de México, México, D.F. 04510 (Mexico); and others

2016-05-15

Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.

Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

International Nuclear Information System (INIS)

Espinosa-Paredes, G.; Prieto-Guerrero, A.; Núñez-Carrera, A.; Vázquez-Rodríguez, A.; Centeno-Pérez, J.; Espinosa-Martínez, E.-G.

2016-01-01

Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.

Numerical analysis of performance of steam reformer of methane reforming hydrogen production system connected with high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Yin Huaqiang; Jiang Shengyao; Zhang Youjie

2007-01-01

Methane conversion rate and hydrogen output are important performance indexes of the steam reformer. The paper presents numerical analysis of performance of the reformer connected with high-temperature gas-cooled reactor HTR-10. Setting helium inlet flow rate fixed, performance of the reformer was examined with different helium inlet temperature, pressure, different process gas temperature, pressure, flow rate, and different steam to carbon ratio. As the range concerned, helium inlet temperature has remarkable influence on the performance, and helium inlet temperature, process gas temperature and pressure have little influence on the performance, and improving process gas flow rate, methane conversion rate decreases and hydrogen output increases, however improving steam to carbon ratio has reverse influence on the performance. (authors)

Conceptual design of a hydrogen production system by DME steam reforming and high-efficiency nuclear reactor technology

International Nuclear Information System (INIS)

Fukushima, Kimichika; Ogawa, Takashi

2003-01-01

Hydrogen is a potential alternative energy source and produced commercially by methane (natural gas) or LPG steam reforming, a process that requires high temperatures, which are produced by burning fossil fuels. However, since this process emits large amounts of CO 2 , replacement of the combustion heat source with a nuclear heat source for 773-1173 K processes has been proposed in order to eliminate these CO 2 emissions. This paper proposes a novel method of low-temperature nuclear hydrogen production by reforming dimethyl ether (DME) with steam produced by a low-temperature nuclear reactor at about 573 K. The authors identified conditions that provide high hydrogen production fraction at low pressure and temperatures of about 523-573 K. By setting this low-temperature hydrogen production process at about 573K upstream from a turbine, it was found theoretically that the total energy utilization efficiency is about 50% and very high. By setting a turbine upstream of the hydrogen production plant, an overall efficiency of is 75% for an FBR and 76% for a supercritical-water cooled power reactor (SCPR). (author)

Cooling facility for reactor container

Energy Technology Data Exchange (ETDEWEB)

Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Kinoshita, Shoichiro

1996-05-31

A suction port of a condensator to a condensate pipe is connected to a main steam pipe, a discharge port of a incondensible gas exhaustion pipe is connected from an inlet header of the condensator to a main steam pipe by way of a valve, and an exhaustion port of the incondensible gas exhaustion pipe is connected from an exit header of the condensator to a pressure suppression pool by way of a valve. In addition, a condensate return pipe is connected from the exit header of the condensator to the pressure vessel by way of a value. When the reactor is isolated, steams are flown from the pressure vessel to a condensator by way of a main steam pipe. In this case, since incondensible gas is not present, the flow rate of inflown steams is great, the condensate heat conductivity is great and temperature difference between the inside and the outside of the pipes is great, the amount of heat released out of the container is increased. The value of the condensate return pipe is opened, condensates are injected to the pressure vessel. Upon occurrence of an accident, steams and incondensible gases are mixed and flown from the suction pipe of the condensator into the condensator, and noncondensed steams are discharged to a pressure suppression pool by the pressure difference between the inside of the condensate pipe and the inside of the pressure suppression chamber. (N.H.)

Cooling facility for reactor container

International Nuclear Information System (INIS)

Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Kinoshita, Shoichiro.

1996-01-01

A suction port of a condensator to a condensate pipe is connected to a main steam pipe, a discharge port of a incondensible gas exhaustion pipe is connected from an inlet header of the condensator to a main steam pipe by way of a valve, and an exhaustion port of the incondensible gas exhaustion pipe is connected from an exit header of the condensator to a pressure suppression pool by way of a valve. In addition, a condensate return pipe is connected from the exit header of the condensator to the pressure vessel by way of a value. When the reactor is isolated, steams are flown from the pressure vessel to a condensator by way of a main steam pipe. In this case, since incondensible gas is not present, the flow rate of inflown steams is great, the condensate heat conductivity is great and temperature difference between the inside and the outside of the pipes is great, the amount of heat released out of the container is increased. The value of the condensate return pipe is opened, condensates are injected to the pressure vessel. Upon occurrence of an accident, steams and incondensible gases are mixed and flown from the suction pipe of the condensator into the condensator, and noncondensed steams are discharged to a pressure suppression pool by the pressure difference between the inside of the condensate pipe and the inside of the pressure suppression chamber. (N.H.)

Steam generator life management

International Nuclear Information System (INIS)

Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

1998-01-01

Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermal hydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. This report shows how recent advances in cleaning technology are integrated into a life management strategy, discusses downcomer flow measurement as a means of monitoring steam generator condition, and describes recent advances in hideout return as a life management tool. The research and development program, as well as operating experience, has identified

French steam generator design developments

International Nuclear Information System (INIS)

Ginier, R.; Campan, J.L.; Pontier, M.; Leridon, A.; Remond, A.; Castello, G.; Holcblat, A.; Paurobally, H.

1986-01-01

From the outset of the French nuclear power program, a significant R and D effort has been invested in improvement of the design and operation of Pressurized Water Reactors including a special committment to improving steam generators. The steam generator enhancement program has spawned a wide variety of specific R and D resources, e.g., low temperature hydraulic models for investigation of areas with single-phase flow, and freon-filled models for simulation of areas of steam generators experiencing two-phase flow (tube bundles and moisture separators). For the moisture separators, a large scale research program using freon-filled models and highly sophisticated instrumentation was used. Tests at reactor sites during startup of both 900 MWe and 1300 MWe have been used to validate the assumptions made on the basis of loop tests. These tests also demonstrated the validity of using freon to simulate two-phase flow conditions. The wealth of knowledge accumulated by the steam generator R and D program has been used to develop a new design of steam generators for the N4 plants. The current R and D effort is aimed at qualifying the N4 steam generator model and developing more comprehensive models. One prong of the R and D effort is the Megeve program. Megeve is a 25 MW steam generator which simulates operating conditions of the N4 model. The other prong is Clotaire, a freon-filled steam generator model which will be used to qualify thermal/hydraulic design codes used for multidimensional calculations for design of tube bundles

CANDU steam generator life management

International Nuclear Information System (INIS)

Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

1998-01-01

Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. The research and development program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factory. (author)

Thermal performances of molten salt steam generator

International Nuclear Information System (INIS)

Yuan, Yibo; He, Canming; Lu, Jianfeng; Ding, Jing

2016-01-01

Highlights: • Thermal performances of molten salt steam generator were experimentally studied. • Overall heat transfer coefficient reached maximum with optimal molten salt flow rate. • Energy efficiency first rose and then decreased with salt flow rate and temperature. • Optimal molten salt flow rate and temperature existed for good thermal performance. • High inlet water temperature benefited steam generating rate and energy efficiency. - Abstract: Molten salt steam generator is the key technology for thermal energy conversion from high temperature molten salt to steam, and it is used in solar thermal power station and molten salt reactor. A shell and tube type molten salt steam generator was set up, and its thermal performance and heat transfer mechanism were studied. As a coupling heat transfer process, molten salt steam generation is mainly affected by molten salt convective heat transfer and boiling heat transfer, while its energy efficiency is also affected by the heat loss. As molten salt temperature increased, the energy efficiency first rose with the increase of heat flow absorbed by water/steam, and then slightly decreased for large heat loss as the absorbed heat flow still rising. At very high molten salt temperature, the absorbed heat flow decreased as boiling heat transfer coefficient dropping, and then the energy efficiency quickly dropped. As the inlet water temperature increased, the boiling region in the steam generator remarkably expanded, and then the steam generation rate and energy efficiency both rose with the overall heat transfer coefficient increasing. As the molten salt flow rate increased, the wall temperature rose and the boiling heat transfer coefficient first increased and then decreased according to the boiling curve, so the overall heat transfer coefficient first increased and then decreased, and then the steam generation rate and energy efficiency of steam generator both had maxima.

Assessment of steam explosion impact on KNGR plant

Energy Technology Data Exchange (ETDEWEB)

Hwang, Moon Kyu; Park, Soo Yong; Park, Ik Kyu

1999-03-01

In present day light water reactors, if complete and prolonged failure of normal and emergency coolant flow occurs, fission product decay heat could cause melting of the reactor fuel. If the molten fuel mass accumulates it may relocate into reactor lower plenum and if the lower head fails it may eventually be brought into the reactor cavity. In such course of core melt relocation, the opportunity for fuel-coolant interactions (FCI) arises as the core melt relocates into water pool in reactor vessel as well as in reactor cavity and also, as a consequence of implementing accident management strategies involving water addition to a degraded or molten core. This report presents the methodologies and their results for assessment of steam explosion impact on KNGR plant integrity. Both in-vessel and ex-vessel phenomena are addressed. For in-vessel steam explosion, TRACER-II code is used for assessment of pressure load, while bounding calculations are applied for ex-vessel analysis. Analysis shows that the integrity of reactor pressure vessel lower head is preserved during the in-vessel event and the probability that the containment integrity is challenged is very low, even when ex-vessel steam explosion is allowed due to reactor vessel failure. (Author). 15 refs., 2 tabs., 4 figs.

Assessment of steam explosion impact on KNGR plant

International Nuclear Information System (INIS)

Hwang, Moon Kyu; Park, Soo Yong; Park, Ik Kyu

1999-03-01

In present day light water reactors, if complete and prolonged failure of normal and emergency coolant flow occurs, fission product decay heat could cause melting of the reactor fuel. If the molten fuel mass accumulates it may relocate into reactor lower plenum and if the lower head fails it may eventually be brought into the reactor cavity. In such course of core melt relocation, the opportunity for fuel-coolant interactions (FCI) arises as the core melt relocates into water pool in reactor vessel as well as in reactor cavity and also, as a consequence of implementing accident management strategies involving water addition to a degraded or molten core. This report presents the methodologies and their results for assessment of steam explosion impact on KNGR plant integrity. Both in-vessel and ex-vessel phenomena are addressed. For in-vessel steam explosion, TRACER-II code is used for assessment of pressure load, while bounding calculations are applied for ex-vessel analysis. Analysis shows that the integrity of reactor pressure vessel lower head is preserved during the in-vessel event and the probability that the containment integrity is challenged is very low, even when ex-vessel steam explosion is allowed due to reactor vessel failure. (Author). 15 refs., 2 tabs., 4 figs

Steam generators

International Nuclear Information System (INIS)

Hayden, R.L.J.

1979-01-01

Steam generators for nuclear reactors are designed so that deposition of solids on the surface of the inlet side of the tubesheet or the inlet header with the consequent danger of corrosion and eventual tube failure is obviated or substantially reduced. (U.K.)

Pulsed high-pressure (PHP) drain-down of steam generating system

International Nuclear Information System (INIS)

Petrusek, R.A.

1991-01-01

This patent describes an improved method of draining down contained reactor-coolant water from the inverted vertical U-tubes of at least one vertical-type steam generator in which the upper inverted U-shaped ends of the tubes are closed and the lower ends thereof are open, the steam generator having a channel head at its lower end including a vertical dividing wall defining a primary water inlet side and a primary water outlet side of the generator, the steam generator having chemical volume control system means and residual heat removal system means, and the steam generator being part of a nuclear-powered steam generating system wherein the reactor-coolant water is normally circulated from and back into the reactor via a loop comprising the steam generator and inlet and outlet conduits connected to the lower end of the steam generator, and the reactor being in communication with pressurizer means and comprising the steps of introducing a gas which is inert to the system and which is under pressure above atmospheric pressure into at least one of the downwardly facing open ends of each of the U-tubes from below the tubesheet in which the open ends of the U-tubes are mounted adjacent the lower end of the steam generator while permitting the water to flow out from the open ends of the U-tubes, the improvement in combination therewith for substantially increasing the effectiveness and efficiency of such water removal from the tubes. It includes determining the parameters effecting a first average volumetric rate of removal for a predetermined period of time, infra, of the reactor-coolant water from the inverted vertical U-tubes, the specific unit for the first average volumetric rate expressing properties identical with the properties expressed in a second average volumetric rate maintained in a later mentioned step

Design codes for gas cooled reactor components

International Nuclear Information System (INIS)

1990-12-01

High-temperature gas-cooled reactor (HTGR) plants have been under development for about 30 years and experimental and prototype plants have been operated. The main line of development has been electricity generation based on the steam cycle. In addition the potential for high primary coolant temperature has resulted in research and development programmes for advanced applications including the direct cycle gas turbine and process heat applications. In order to compare results of the design techniques of various countries for high temperature reactor components, the IAEA established a Co-ordinated Research Programme (CRP) on Design Codes for Gas-Cooled Reactor Components. The Federal Republic of Germany, Japan, Switzerland and the USSR participated in this Co-ordinated Research Programme. Within the frame of this CRP a benchmark problem was established for the design of the hot steam header of the steam generator of an HTGR for electricity generation. This report presents the results of that effort. The publication also contains 5 reports presented by the participants. A separate abstract was prepared for each of these reports. Refs, figs and tabs

Steam generator for PWR type reactor

International Nuclear Information System (INIS)

Baba, Iwao; Hiyama, Nobuyuki.

1994-01-01

A steam generator of the present invention comprises a primary coolant chamber having primary coolants circulating therein, a secondary coolants chamber having secondary coolants and steams circulating therein, which are isolated from each other by a partition wall, and heat pipes disposed being passed through the partition wall. The heat pipes are disposed having an evaporation portion in the primary coolants chamber, a condensation portion in the secondary coolants chamber, and an intermediate heat insulating portion in the partition wall. Since the primary coolants containing radioactivity and the secondary coolants not containing radioactivity does not transfer heat directly by a heat transfer wall, a leakage accident of radioactivity to the secondary coolants can be prevented. Moreover, since the heat pipes are used, a great amount of heat can be transferred by a slight temperature difference by using steams of the heat transfer medium itself, latent heat due to coagulation, and capillary phenomenon. Since neither transferring power nor pumps are required, heat of the primary coolants can effectively be transferred to the secondary coolants. (N.H.)

Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

International Nuclear Information System (INIS)

Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

1997-06-01

This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked ampersand influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs

Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

Energy Technology Data Exchange (ETDEWEB)

Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

1997-06-01

This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

Studies on high temperature research reactor in China

Energy Technology Data Exchange (ETDEWEB)

Xu Yuanhui; Zuo Kanfen [Institute of Nuclear Energy Technology, Tsinghua Univ., Beijing (China)

1999-08-01

China recognises the advantages of Modular HTGRs and has chosen Modular HTGRs as one of advanced reactors to be developed for the further intensive utilisation of nuclear power in the next century. In energy supply systems of the next century, HTGR is supposed to serve: 1. as supplement to water-cooled reactors for electricity generation and 2. as environmentally friendly heat source providing process heat at different temperatures for various applications like heavy oil recovery, coal gasification and liquefaction, etc.. The 10 MW High Temperature Gas-cooled Reactor (HTR-10) is a major project in the energy sector of the Chinese National High Technology Programme as the first step of development of Modular HTGRs in China. Its main objectives are: 1. to acquire know-how in the design, construction and operation of HTGRs, 2. to establish an irradiation and experimental facility, 3. to demonstrate the inherent safety features of Modular HTGR, 4. to test electricity and heat co-generation and closed cycle gas turbine technology and 5. to do research and development work on the nuclear process heat application. Now the HTR-10 is being constructed at the site of Institute of Nuclear Energy Technology (INET). The HTR-10 project is to be carried out in two phases. In the first phase, the reactor with an coolant outlet temperature of 700degC will be coupled with a steam generator providing steam for a steam turbine cycle which works on an electricity and heat co-generation basis. In the second phase, the reactor coolant outlet temperature is planned to be raised to 900degC. As gas turbine cycle and a steam reformer will be coupled to the reactor in addition to the steam turbine cycle. (author)

Studies on high temperature research reactor in China

International Nuclear Information System (INIS)

Xu Yuanhui; Zuo Kanfen

1999-01-01

China recognises the advantages of Modular HTGRs and has chosen Modular HTGRs as one of advanced reactors to be developed for the further intensive utilisation of nuclear power in the next century. In energy supply systems of the next century, HTGR is supposed to serve: 1. as supplement to water-cooled reactors for electricity generation and 2. as environmentally friendly heat source providing process heat at different temperatures for various applications like heavy oil recovery, coal gasification and liquefaction, etc.. The 10 MW High Temperature Gas-cooled Reactor (HTR-10) is a major project in the energy sector of the Chinese National High Technology Programme as the first step of development of Modular HTGRs in China. Its main objectives are: 1. to acquire know-how in the design, construction and operation of HTGRs, 2. to establish an irradiation and experimental facility, 3. to demonstrate the inherent safety features of Modular HTGR, 4. to test electricity and heat co-generation and closed cycle gas turbine technology and 5. to do research and development work on the nuclear process heat application. Now the HTR-10 is being constructed at the site of Institute of Nuclear Energy Technology (INET). The HTR-10 project is to be carried out in two phases. In the first phase, the reactor with an coolant outlet temperature of 700degC will be coupled with a steam generator providing steam for a steam turbine cycle which works on an electricity and heat co-generation basis. In the second phase, the reactor coolant outlet temperature is planned to be raised to 900degC. As gas turbine cycle and a steam reformer will be coupled to the reactor in addition to the steam turbine cycle. (author)

Discussion on amount of water ingress mass in steam generator heat-exchange tube rupture accident of high- temperature gas-cooled reactor

International Nuclear Information System (INIS)

Wang Yan; Zheng Yanhua; Shi Lei; Li Fu; Sun Ximing

2009-01-01

The steam generator heat-exchange tube rupture (SGTR) accident which will result in the water ingress to the primary circuit of reactor is an important and particular accident for high-temperature gas-cooled reactor (HTGR). The analysis of the water ingress accident is significant for verifying the inherent safety characteristics of HTGR. The amount of water ingress mass is one of the decisive factors for the seriousness of the accident consequence. The 250 MW Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) designed by Institute of Nuclear and New Energy Technology of Tsinghua University was selected as an example of analysis. The analysis results show that the amount of water ingress mass is not only affected directly with the broken position and the broken area of the tubes, but also related with the diameter of draining piping and restrictor, draining control valve, action setting of emptier system. With reasonable parameters chosen, the water in steam generator could be drained effectively, so it will prevent the primary circuit of reactor from water ingress in large quantity and reduce the radioactive isotopes ingress to the secondary circuit. (authors)

Direct injection of superheated steam for continuous hydrolysis reaction

KAUST Repository

Wang, Weicheng

2012-09-01

The primary intent for previous continuous hydrolysis studies was to minimize the reaction temperature and reaction time. In this work, hydrolysis is the first step of a proprietary chemical process to convert lipids to sustainable, drop-in replacements for petroleum based fuels. To improve the economics of the process, attention is now focused on optimizing the energy efficiency of the process, maximizing the reaction rate, and improving the recovery of the glycerol by-product. A laboratory-scale reactor system has been designed and built with this goal in mind.Sweet water (water with glycerol from the hydrolysis reaction) is routed to a distillation column and heated above the boiling point of water at the reaction pressure. The steam pressure allows the steam to return to the reactor without pumping. Direct injection of steam into the hydrolysis reactor is shown to provide favorable equilibrium conditions resulting in a high quality of FFA product and rapid reaction rate, even without preheating the inlet water and oil and with lower reactor temperatures and lower fresh water demand. The high enthalpy of the steam provides energy for the hydrolysis reaction. Steam injection offers enhanced conditions for continuous hydrolysis of triglycerides to high-purity streams of FFA and glycerol. © 2012 Elsevier B.V.

STEAM ENHANCED REMEDIATION RESEARCH FOR DNAPL IN FRACTURED ROCK, LORING AIR FORCE BASE, LIMESTONE, MAINE

Science.gov (United States)

This report details a research project on Steam Enhanced Remediation (SER) for the recovery of volatile organic compounds from fractured limestone that was carried out at the Quarry at the former Loring Air Force Base in Limestone, Maine. This project was carried out by USEPA, Ma...

Simulation of the Thermal Hydraulic Processes in the Horizontal Steam Generator with the Use of the Different Interfacial Friction Correlations

International Nuclear Information System (INIS)

Melikhov, V.; Melikhov, O.; Parfenov, Y.; Nerovnov, A.

2011-01-01

The horizontal steam generator (SG) is one of specific features of Russian-type pressurized water reactors (VVERs). The main advantages of horizontal steam generator are connected with low steam loads on evaporation surface, simple separation scheme and high circulation ratio. The complex three-dimensional steam-water flows in the steam generator vessel influence significantly the processes of the steam separation, distribution, and deposition of the soluble and non soluble impurities and determine the efficiency and reliability of the steam generator operation. The 3D code for simulation of the three-dimensional steam-water flows in the steam generator could be effective tool for design and optimization of the horizontal steam generator. The results of the code calculations are determined mainly by the set of the correlations describing interaction of the steam-water mixture with the inner constructions of the SG and interfacial friction. The results obtained by 3D code STEG with the usage of the different interfacial friction correlations are presented and discussed in the paper. These results are compared with the experimental ones obtained at the experimental test facility PGV-1500 constructed for investigation of the processes in the horizontal steam generator

Reevaluation of steam generator level trip set point

Energy Technology Data Exchange (ETDEWEB)

Shim, Yoon Sub; Soh, Dong Sub; Kim, Sung Oh; Jung, Se Won; Sung, Kang Sik; Lee, Joon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

The reactor trip by the low level of steam generator water accounts for a substantial portion of reactor scrams in a nuclear plant and the feasibility of modification of the steam generator water level trip system of YGN 1/2 was evaluated in this study. The study revealed removal of the reactor trip function from the SG water level trip system is not possible because of plant safety but relaxation of the trip set point by 9 % is feasible. The set point relaxation requires drilling of new holes for level measurement to operating steam generators. Characteristics of negative neutron flux rate trip and reactor trip were also reviewed as an additional work. Since the purpose of the trip system modification for reduction of a reactor scram frequency is not to satisfy legal requirements but to improve plant performance and the modification yields positive and negative aspects, the decision of actual modification needs to be made based on the results of this study and also the policy of a plant owner. 37 figs, 6 tabs, 14 refs. (Author).

Inverted Steam Generators for Sodium Cooled Fast Reactors

International Nuclear Information System (INIS)

Matal, Oldřich; Šimo, Tomáš; Matal, Oldřich Jr.

2013-01-01

Conclusions: Two inverted steam generators of the Czech industry provenience have still been in successful operation with no water into sodium leaks at BOR 60 (RIAR Dimitrovgrad, Russian Federation). Micromodular inverted steam generator (MMISG) since 1981 and modular inverted steam generator (MISG) since 1991. In the framework of the CP ESFR project predesign studies of 100 MW (thermal) ISG modules were performed with the consideration of MMISG and MISG design, operational and safety benefits and experience. Development of material and technology for sodium heated steam generators components reflecting contemporary domestic industrial conditions in the Czech Republic was restarted in the years 2003 to 2004 and supported in the years 2008 to 2011 by the European CP ESFR project and by the Ministry of Industry and Trade of the Czech Republic

Design and construction features of steam generators at a nuclear power station

International Nuclear Information System (INIS)

Chakrabarti, A.K.; Gupta, K.N.; Bapat, C.N.; Sharma, V.K.

1996-01-01

The Indian nuclear power programme is based on Pressurised Heavy Water Reactors (PHWRs) using natural uranium as fuel and heavy water as reactor coolant as well as moderator. The nuclear heat is generated in the fuel located in the pressure tubes. Pressurised heavy water in the primary heat transport (PHT) system is circulated through the tubes which picks up the heat from the fuel and transfers it to ordinary water in steam generators (SGs) to produce steam. The steam is used for providing power to the turbine. The steam generator is a critical equipment in the nuclear steam supply system (NSSS) of a nuclear reactor. SG tube surface area constitute about 80% of total primary circuit surface area. A typical value in a 220 MWe reactor is 9000 m 2 which can release considerable amount of corrosion products unless very low corrosion rates are achieved by proper design, material selection and water chemistry control. Design and construction features of SGs are given. 1 tab

Systems Analysis of a Fast Steam-Cooled Reactor of 1000 MW(E)

Energy Technology Data Exchange (ETDEWEB)

Smidt, D.; Frisch, W.; Hofmann, F.; Moers, H.; Schramm, K.; Spilker, H. [Institut fuer Reaktorentwicklung, Kernforschungszentrum, Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Kiefhaber, E. [Institut fuer Neutronenphysik und Reaktortechnik Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)

1968-05-15

The Karlsruhe design of a steam-cooled fast reactor (Dl) has been the subject of a systems analysis. Here the dependence of fuel inventory, breeding ratio, rating, core geometry and plant efficiency on coolant pressure, and coolant temperature has been studied for two different rod powers. The effect of artificial surface roughness has been investigated. For some configurations the resulting fuel-cycle and capital costs have been determined and discussed. The main influence results from pressure. The lower pressure allows for higher breeding ratios, but lower efficiencies and vice versa. From this the fuel-cycle costs show an optimum at around 150 atm abs. The capital costs on the other side decrease with pressure. The over-all optimum of the power generating costs for the presently studied parameter range is at about 170 atm abs., a coolant outlet temperature of 540 Degree-Sign C and a rod power of 420 W/cm. Artificial roughness (boundary layer type) leads for a required system pressure and outlet temperature to a larger coolant volume fraction and, therefore, to reduced breeding ratios but higher efficiencies. As another part of the work some stability characteristics of the cores were studied. The dependence of the core stability on the varied parameters is shown. (author)

Reactor type choice and characteristics for a small nuclear heat and electricity co-generation plant

International Nuclear Information System (INIS)

Liu Kukui; Li Manchang; Tang Chuanbao

1997-01-01

In China heat supply consumes more than 70 percent of the primary energy resource, which makes for heavy traffic and transportation and produces a lot of polluting materials such as NO x , SO x and CO 2 because of use of the fossil fuel. The utilization of nuclear power into the heat and electricity co-generation plant contributes to the global environmental protection. The basic concept of the nuclear system is an integral type reactor with three circuits. The primary circuit equipment is enclosed in and linked up directly with reactor vessel. The third circuit produces steam for heat and electricity supply. This paper presents basic requirements, reactor type choice, design characteristics, economy for a nuclear co-generation plant and its future application. The choice of the main parameters and the main technological process is the key problem of the nuclear plant design. To make this paper clearer, take for example a double-reactor plant of 450 x 2MW thermal power. There are two sorts of main technological processes. One is a water-water-steam process. Another is water-steam-steam process. Compared the two sorts, the design which adopted the water-water-steam technological process has much more advantage. The system is simplified, the operation reliability is increased, the primary pressure reduces a lot, the temperature difference between the secondary and the third circuits becomes larger, so the size and capacity of the main components will be smaller, the scale and the cost of the building will be cut down. In this design, the secondary circuit pressure is the highest among that of the three circuits. So the primary circuit radioactivity can not leak into the third circuit in case of accidents. (author)

Engineering design of a direct-cycle steam-generating blanket for a long-pulse fusion reactor

International Nuclear Information System (INIS)

Cort, G.E.; Hagenson, R.L.; Teasdale, R.W.; Fox, W.E.; Soran, P.D.; Cullingford, H.S.; Bathke, C.G.; Krakowski, R.A.

1979-01-01

A comprehensive neutronics, thermohydraulic, and mechanical design of a tritium-breeding blanket for use by a conceptual long-pulse Reversed-Field Pinch Reactor (RFPR) is described. On the basis of constraints imposed by cost and the desire to use existing technology, a direct-cycle steam system and stainless-steel construction were used. For reasons of plasma stability, the RFPR blanket supports a 20-mm-thick copper first wall. Located behind the 1.5-m-radius first wall is a 0.50-m-thick stainless-steel blanket containing a granular bed of Li 2 O through which flows low-pressure helium (0.1 MPa) for tritium extraction. Water/steam tubes radially penetrate this packed bed. The large thermal capacity and low thermal diffusivity of the Li 2 O blanket are sufficient to maintain a nearly constant temperature during the approx. 25-s burn period

OECD/NEZ Main Steam Line Break Benchmark Problem Exercise I Simulation Using the SPACE Code with the Point Kinetics Model

International Nuclear Information System (INIS)

Kim, Yohan; Kim, Seyun; Ha, Sangjun

2014-01-01

The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Nuclear Hydro and Nuclear Power Co. (KHNP) through collaborative works with other Korean nuclear industries. The SPACE is a best-estimated two-phase three-field thermal-hydraulic analysis code to analyze the safety and performance of pressurized water reactors (PWRs). The SPACE code has sufficient features to replace outdated vendor supplied codes and to be used for the safety analysis of operating PWRs and the design of advanced reactors. As a result of the second phase of the development, the 2.14 version of the code was released through the successive various V and V works. The topical reports on the code and related safety analysis methodologies have been prepared for license works. In this study, the OECD/NEA Main Steam Line Break (MSLB) Benchmark Problem Exercise I was simulated as a V and V work. The results were compared with those of the participants in the benchmark project. The OECD/NEA MSLB Benchmark Problem Exercise I was simulated using the SPACE code. The results were compared with those of the participants in the benchmark project. Through the simulation, it was concluded that the SPACE code can effectively simulate PWR MSLB accidents

Examination of failed studs from No. 2 steam generator at the Maine Yankee Nuclear Power Station

International Nuclear Information System (INIS)

Czajkowski, C.

1983-02-01

Three studs removed from service on the primary manway cover from steam generator No. 2 of the Maine Yankee station were sent to Brookhaven National Laboratory (BNL) for examination. The examination consisted of visual/dye penetrant examination, optical metallography and Scanning Electron Microscopy/Energy Dispersive Spectroscopy (SEM/EDS) evaluation. One bolt was through cracked and its fracture face was generally transgranular in nature with numerous secondary intergranular cracks. The report concludes that the environmenally assisted cracking of the stud was due to the interaction of the various lubricants used with steam leaks associated with this manway cover

Steam producing plant concept of 4S for oil sand extraction

International Nuclear Information System (INIS)

Matsuyama, Shinichiro; Nishiguchi, Youhei; Sakashita, Yoshiaki; Kasuga, Shoji; Kawashima, Masatoshi

2009-01-01

Plant concept of small fast reactor '4S' applying to continuous steam production for recovery of crude oil from oil sands was investigated. Assuming typical steam assisted gravity drainage (SAGD) plant whose production scale is 120,000 barrels per day of a crude oil, concept of nuclear steam supply system consisting of eight reactor modules for steam production and three reactor modules for electric generation of the 4S with a thermal rating of 135 MWt was established without any essential or significant design change from the preceding 4S with a thermal rating of 30 MWt. The 4S, provided for an oil sand extraction, will reduce greenhouse gas emission significantly, and has not much burden for development and licensing and has economic competitiveness. (author)

BWR Steam Dryer Alternating Stress Assessment Procedures

Energy Technology Data Exchange (ETDEWEB)

Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

2016-12-01

This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

PWR type reactor plant

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi.

1993-01-01

A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

Strain measurement on a compact nuclear reactor steam generator

International Nuclear Information System (INIS)

Scaldaferri, Denis Henrique Bianchi; Gomes, Paulo de Tarso Vida; Mansur, Tanius Rodrigues; Pozzo, Renato del; Mola, Jairo

2011-01-01

This work presents the strain measurement procedures applied to a compact nuclear reactor steam generator, during a hydrostatic test, using strain gage technology. The test was divided in two steps: primary side test and secondary side test. In the primary side test twelve points for strain measurement using rectangular rosettes, three points (two external and one internal) for temperature measurement using special strain gages and one point for pressure measurement using a pressure transducer were monitored. In the secondary side test 18 points for strain measurement using rectangular rosettes, four points (two external and two internal) for temperature measurement using special strain gages and one point for pressure measurement using a pressure transducer were monitored. The measurement points on both internal and external pressurizer walls were established from pre-calculated stress distribution by means of numerical approach (finite elements modeling). Strain values using a quarter Wheatstone bridge circuit were obtained. Stress values, from experimental strain were determined, and to numerical calculation results were compared. (author)

Dynamic simulation of pure hydrogen production via ethanol steam reforming in a catalytic membrane reactor

International Nuclear Information System (INIS)

Hedayati, Ali; Le Corre, Olivier; Lacarrière, Bruno; Llorca, Jordi

2016-01-01

Ethanol steam reforming (ESR) was performed over Pd-Rh/CeO 2 catalyst in a catalytic membrane reactor (CMR) as a reformer unit for production of fuel cell grade pure hydrogen. Experiments were performed at 923 K, 6–10 bar, and fuel flow rates of 50–200 μl/min using a mixture of ethanol and distilled water with steam to carbon ratio of 3. A static model for the catalytic zone was derived from the Arrhenius law to calculate the total molar production rates of ESR products, i.e. CO, CO 2 , CH 4 , H 2 , and H 2 O in the catalytic zone of the CMR (coefficient of determination R 2  = 0.993). The pure hydrogen production rate at steady state conditions was modeled by means of a static model based on the Sieverts' law. Finally, a dynamic model was developed under ideal gas law assumptions to simulate the dynamics of pure hydrogen production rate in the case of the fuel flow rate or the operating pressure set point adjustment (transient state) at isothermal conditions. The simulation of fuel flow rate change dynamics was more essential compared to the pressure change one, as the system responded much faster to such an adjustment. The results of the dynamic simulation fitted very well to the experimental values at P = 7–10 bar, which proved the robustness of the simulation based on the Sieverts' law. The simulation presented in this work is similar to the hydrogen flow rate adjustments needed to set the electrical load of a fuel cell, when fed online by the pure hydrogen generating reformer studied. - Highlights: • Ethanol steam reforming (ESR) experiments were performed in a Pd-Ag membrane reactor. • The model of the catalytic zone of the reactor was derived from the Arrhenius law. • The permeation zone (membrane) was modeled based on the Sieverts' law. • The Sieverts' law model showed good results for the range of P = 7–10 bar. • Pressure and fuel flow rate adjustments were considered for dynamic simulation.

Status of and prospects for gas-cooled reactors

International Nuclear Information System (INIS)

1984-01-01

The IAEA International Working Group on Gas-Cooled Reactors (IWGGCR) (see Annex I), which was established in 1978, recommended to the Agency that a report be prepared in order to provide an up-to-date summary of gas-cooled reactor technology. The present Technical Report is based mainly on submissions of Member Countries of the IWGGCR and consists of four main sections. Beside some general information about the gas-cooled reactor line, section 1 contains a description of the incentives for the development and deployment of gas-cooled reactors in various Agency Member States. These include both electricity generation and process steam and process heat production for various branches of industry. The historical development of gas-cooled reactors is reviewed in section 2. In this section information is provided on how, when and why gas-cooled reactors have been developed in various Agency Member States and, in addition, a detailed description of the different gas-cooled reactor lines is presented. Section 3 contains information about the technical status of gas-cooled reactors and their applications. Gas-cooled reactors that are under design or construction or in operation are listed and shortly described, together with an outlook for future reactor designs. In this section the various applications for gas-cooled reactors are described in detail. These include both electricity generation and process steam and process heat production. The last section (section 4) is entitled ''Special features of gas-cooled reactors'' and contains information about the technical performance, fuel utilization, safety characteristics and environmental impact, such as radiation exposure and heat rejection

Status of Siemens steam generator design and measures to assure continuous long-term reliable operation

International Nuclear Information System (INIS)

Hoch, G.

1999-01-01

Operating pressurized water reactors with U-tube steam generators have encountered difficulties with either one or a combination of inadequate material selection, poor design or manufacturing and an insufficient water chemistry control which resulted in excessive tube degradation. In contrast to the above mentioned problems, steam generators from Siemens/KWU are proving by operating experience that all measures undertaken at the design stage as well as during the operating and maintenance phase were effective enough to counteract any tube corrosion phenomena or other steam generator related problem. An Integrated Service Concept has been developed, applied and wherever necessary improved in order to ensure reliable steam generator operation. The performance of the steam generators is updated continuously, evaluated and implemented in lifetime databases. The main indicator for steam generator integrity are the results of the eddy current testing of the steam generator tubes. Tubes with indications are rated with lifetime threshold values and if necessary plugged, based on individual assessment criteria.(author)

Nuclear reactor containing facility

International Nuclear Information System (INIS)

Hidaka, Masataka; Murase, Michio.

1994-01-01

In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

Two Phase Flow Stability in the HTR-10 Steam Generator

Institute of Scientific and Technical Information of China (English)

居怀明; 左开芬; 刘志勇; 徐元辉

2001-01-01

A 10 MW High Temperature Gas Cooled Reactor (HTR-10) designed bythe Institute of Nuclear Energy Technology (INET) is now being constructed. The steam generator (SG) in the HTR-10 is one of the most important components for reactor safety. The thermal-hydraulic performance of the SG was investigated. A full scale HTR-10 Steam Generator Two Tube Engineering Model Test Facility (SGTM-10) was installed and tested at INET. This paper describes the SGTM-10 thermal hydraulic experimental system in detail. The SGTM-10 simulates the actual thermal and structural parameters of the HTR-10. The SGTM-10 includes three separated loops: the primary helium loop, the secondary water loop, and the tertiary cooling water loop. Two parallel tubes are arranged in the test assembly. The main experimental equipment is shown in the paper. Expermental results are given illustrating the effects of the outlet pressures, the heating power, and the inlet subcooling.

Control device for steam turbine

International Nuclear Information System (INIS)

Hoshi, Hiroyuki.

1993-01-01

A power load imbalance detection circuit detects a power load imbalance when a load variation coefficient is large and output-load deviation is great. Then, it self-holds and causes a timer to start counting up and releases the self-holding after the elapse of a certain period of time. Upon load separation caused by system accidents, the power load imbalance detection circuit operates along with the increase of turbine rpm, to operate the control valve abrupt closing circuit and a bypassing value abrupt opening circuit. Then, self-holding of the power load imbalance detection circuit is released and, subsequently, a steam control value and a bypass valve are controlled by a control valve flow rate demand signal and a bypass flow rate demand signal determined by an entire main steam flow rate signal and a speed/load control signal. Accordingly, the turbine rpm is settled to about a rated rpm. This enables to avoid reactor shutdown upon occurrence of load interruption. (I.N.)

Reactor water level control device

International Nuclear Information System (INIS)

Utagawa, Kazuyuki.

1993-01-01

A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

Steam generator replacement at Surry Power Station

International Nuclear Information System (INIS)

McKay, H.S.

1982-01-01

The purposes of the steam generator repair program at Surry Power Station were to repair the tube degradation caused by corrosion-related phenomena and to restore the integrity of the steam generators to a level equivalent to new equipment. The repair program consisted of (1) replacing the existing lower-shell assemblies with new ones and (2) adding new moisture separation equipment to the upper-shell assemblies. These tasks required that several pieces of reactor coolant piping, feedwater piping, main steam piping, and the steam generator be cut and refurbished for reinstallation after the new lower shell was in place. The safety implications and other potential effects of the repair program both during the repair work and after the unit was returned to power were part of the design basis of the repair program. The repair program has been completed on Unit 2 without any adverse effects on the health and safety of the general public or to the personnel engaged in the repair work. Before the Unit 1 repair program began, a review of work procedures and field changes for the Unit 2 repair was conducted. Several major changes were made to avoid recurrence of problems and to streamline procedures. Steam generator replacements was completed on June 1, 1981, and the unit is presently in the startup phase of the outrage

Leak before break evaluation for main steam piping system made of SA106 Gr.C

Energy Technology Data Exchange (ETDEWEB)

Yang, Kyoung Mo; Jee, Kye Kwang; Pyo, Chang Ryul; Ra, In Sik [Korea Power Engineering Company, Seoul (Korea, Republic of)

1997-04-01

The basis of the leak before break (LBB) concept is to demonstrate that piping will leak significantly before a double ended guillotine break (DEGB) occurs. This is demonstrated by quantifying and evaluating the leak process and prescribing safe shutdown of the plant on the basis of the monitored leak rate. The application of LBB for power plant design has reduced plant cost while improving plant integrity. Several evaluations employing LBB analysis on system piping based on DEGB design have been completed. However, the application of LBB on main steam (MS) piping, which is LBB applicable piping, has not been performed due to several uncertainties associated with occurrence of steam hammer and dynamic strain aging (DSA). The objective of this paper is to demonstrate the applicability of the LBB design concept to main steam lines manufactured with SA106 Gr.C carbon steel. Based on the material properties, including fracture toughness and tensile properties obtained from the comprehensive material tests for base and weld metals, a parametric study was performed as described in this paper. The PICEP code was used to determine leak size crack (LSC) and the FLET code was used to perform the stability assessment of MS piping. The effects of material properties obtained from tests were evaluated to determine the LBB applicability for the MS piping. It can be shown from this parametric study that the MS piping has a high possibility of design using LBB analysis.

Control of WWER-440 nuclear reactor

International Nuclear Information System (INIS)

Wagner, K.; Drab, F.; Grof, V.

1978-01-01

The V-1 reactor control systems are described. The data acquisition and processing system fulfils four main functions, ie., reactor start-up and power increase to 10% of the rated power, automatic power control within 3% and 110% of the rated power, reactivity compensation, and reactor protection. The automatic control system ensures constant steam pressure maintained with an accuracy of +-0.05 MPa. Reactivity compensation and spatial power distribution is mainly safeguarded by boric acid control. The V-1 reactor protection system has four levels of accident protection depending on the gravity of the failure. The philosophy of automation of the V-1 reactor control and protection system is based on autonomous automatic controlers and on the direct control of the individual sets and technological equipment. In conclusion, development trends are briefly outlined of control and protection systems of light water reactor power plants. (Z.M.)

Reactor System Design

International Nuclear Information System (INIS)

Chi, S. K.; Kim, G. K.; Yeo, J. W.

2006-08-01

SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

Failure analysis of retired steam generator tubings

Energy Technology Data Exchange (ETDEWEB)

Kim, Hong Pyo; Kim, J. S.; Hwang, S. S. and others

2005-04-15

Degradation of steam generator leads to forced outage and extension of outage, which causes increase in repair cost, cost of purchasing replacement power and radiation exposure of workers. Steam generator tube rupture incident occurred in Uljin 4 in 2002, which made public sensitive to nuclear power plant. To keep nuclear energy as a main energy source, integrity of steam generator should be demonstrated. Quantitative relationship between ECT(eddy current test) signal and crack size is needed in assesment of integrity of steam generator in pressurized water reactor. However, it is not fully established for application in industry. Retired steam generator of Kori 1 has many kinds of crack such as circumferential and axial primary water stress corrosion crack and outer diameter stress corrosion crack(ODSCC). So, it can be used in qualifying and improving ECT technology and in condition monitoring assesment for crack detected in ISI(in service inspection). In addition, examination of pulled tube of Kori 1 retired steam generator will give information about effectiveness of non welded sleeving technology which was employed to repair defect tubes and remedial action which was applied to mitigate ODSCC. In this project, hardware such as semi hot lab. for pulled tube examination and modification transportation cask for pulled tube and software such as procedure of transportation of radioactive steam generator tube and non-destructive and destructive examination of pulled tube were established. Non-destructive and destructive examination of pulled tubes from Kori 1 retired steam generator were performed in semi hot lab. Remedial actions applied to Kori 1 retired steam generator, PWSCC trend and bulk water chemistry and crevice chemistry in Kori 1 were evaluated. Electrochemical decontamination technology for pulled tube was developed to reduce radiation exposure and enhance effectiveness of pulled tube examination. Multiparameter algorithm developed at ANL, USA was

Comparison of the updated solutions of the 6th dynamic AER Benchmark - main steam line break in a NPP with WWER-440

International Nuclear Information System (INIS)

Kliem, S.

2003-01-01

The 6 th dynamic AER Benchmark is used for the systematic validation of coupled 3D neutron kinetic/thermal hydraulic system codes. It was defined at The 10 th AER-Symposium. In this benchmark, a hypothetical double ended break of one main steam line at full power in a WWER-440 plant is investigated. The main thermal hydraulic features are the consideration of incomplete coolant mixing in the lower and upper plenum of the reactor pressure vessel and an asymmetric operation of the feed water system. For the tuning of the different nuclear cross section data used by the participants, an isothermal re-criticality temperature was defined. The paper gives an overview on the behaviour of the main thermal hydraulic and neutron kinetic parameters in the provided solutions. The differences in the updated solution in comparison to the previous ones are described. Improvements in the modelling of the transient led to a better agreement of a part of the results while for another part the deviations rose up. The sensitivity of the core power behaviour on the secondary side modelling is discussed in detail (Authors)

Conceptual design of once-through helical steam generator for SMART

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Wan; Kim, J. I.; Kim, J. H. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-12-01

Conceptual design of once-through helical steam generator for the integral reactor SMART is developed. The once-through helical steam generator requires quite different design concepts from the steam generators used in loop type commercial reactors. In this study the design requirements satisfying the operating conditions of the steam generator are derived, and the arrangements and the dimensions of the major parts are determined. By describing the design procedure, the cost of redesign and the costs of developments of similar new steam generators are minimized. The three dimensional models developed make it possible to preview the interferences of the steam generator components and to minimize the possibility of significant design changes in the next design stage by the preliminary strength analysis of the major parts. A methodology for evaluation of flow induced vibration of steam generator tubes has been developed and a preliminary flow induced vibration analysis has been performed. 24 refs., 54 figs., 9 tabs. (Author)

Status of steam gasification of coal by using heat from high-temperature reactors (HTRs)

International Nuclear Information System (INIS)

Schroeter, H.J.; Kirchhoff, R.; Heek, K.H. van; Juentgen, H.; Peters, W.

1984-01-01

Bergbau-Forschung GmbH, Essen, is developing a process for steam gasification of coal by using process heat from high-temperature nuclear reactors (HTRs). The envisaged allothermal gas generator is heated by an internally mounted bundle of heat exchanging tubes through which the gaseous reactor coolant helium flows. Research and development work for this process has been under way for about 11 years. After intensive small-scale investigations the principle of the process was tested in a semi-technical plant with 0.2 t/h coal throughput. In its gasifier a fluidized bed of approximately 1 m 2 cross-section and up to 4 m high is operated at 40 bar. Heat is supplied to the bed from an immersed heat exchanger with helium flowing through it. The gas generator is a cut-out version of the full-scale generator, in which the height of the bed, and the arrangement of the heat-exchanger tubes correspond to the full-scale design. The semi-technical plant has now achieved a total gasification time of about 13,000 hours. Roughly 2000 t of coal have been put through. During recent years the gasification of Federal German coking coal by using a jet-feeding system was demonstrated successfully. The results, confirmed and expanded by material tests for the heat exchanger, engineering and computer models and design studies, have shown the feasibility of nuclear steam gasification of coal. The process described offers the following advantages compared with existing processes: higher efficiency as more gas can be produced from less coal; less emission of pollutants as, instead of a coal-fired boiler, the HTR is used for producing steam and electricity; lower production costs for gas. The next step in the project is a pilot plant of about 2-4 t/h coal throughput, still with non-nuclear heating, to demonstrate the construction and operation of the allothermal gas generator on a representative scale for commercial applications. (author)

Containments for consolidated nuclear steam systems

International Nuclear Information System (INIS)

Jabsen, F.S.

1978-01-01

A containment system for a consolidated nuclear steam system incorporating a nuclear core, steam generator and reactor coolant pumps within a single pressure vessel is described which is designed to provide radiation shielding and pressure suppression. Design details, including those for the dry well and wet well of the containment, are given. (UK)

Horizontal steam generator thermal-hydraulics

Energy Technology Data Exchange (ETDEWEB)

Ubra, O. [SKODA Praha Company, Prague (Czechoslovakia); Doubek, M. [Czech Technical Univ., Prague (Czechoslovakia)

1995-09-01

Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. The 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.