Advanced technologies on steam generators

Energy Technology Data Exchange (ETDEWEB)

Sakata, Kaoru; Nakamura, Yuuki [Mitsubishi Heavy Industry Co., Takasago (Japan); Nakamori, Nobuo; Mizutani, Toshiyuki; Uwagawa, Seiichi; Saito, Itaru [Mitsubishi Heavy Industry Co., Kobe (Japan); Matsuoka, Tsuyoshi [Mitsubishi Heavy Industry Co., Yokohama (Japan)

1997-12-31

The thermal-hydraulic tests for a horizontal steam generator of a next-generation PWR (New PWR-21) were performed. The purpose of these tests is to understand the thermal-hydraulic behavior in the secondary side of horizontal steam generator during the plant normal operation. A test was carried out with cross section slice model simulated the straight tube region. In this paper, the results of the test is reported, and the effect of the horizontal steam generator internals on the thermalhydraulic behavior of the secondary side and the circulation characteristics of the secondary side are discussed. (orig.). 3 refs.

Advanced technologies on steam generators

Energy Technology Data Exchange (ETDEWEB)

Sakata, Kaoru; Nakamura, Yuuki [Mitsubishi Heavy Industry Co., Takasago (Japan); Nakamori, Nobuo; Mizutani, Toshiyuki; Uwagawa, Seiichi; Saito, Itaru [Mitsubishi Heavy Industry Co., Kobe (Japan); Matsuoka, Tsuyoshi [Mitsubishi Heavy Industry Co., Yokohama (Japan)

1998-12-31

The thermal-hydraulic tests for a horizontal steam generator of a next-generation PWR (New PWR-21) were performed. The purpose of these tests is to understand the thermal-hydraulic behavior in the secondary side of horizontal steam generator during the plant normal operation. A test was carried out with cross section slice model simulated the straight tube region. In this paper, the results of the test is reported, and the effect of the horizontal steam generator internals on the thermalhydraulic behavior of the secondary side and the circulation characteristics of the secondary side are discussed. (orig.). 3 refs.

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.

Numerical Study of Thermal Hydraulics for Secondary side of Steam Generator by CUPID/MARS Coupled Simulation

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Ryong; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

As a thermal-hydraulic behavior in the secondary side of steam generator such as two-phase boiling flow, flow-induce vibration of U-tubes is quite complicated, the importance to numerically investigate the flow behavior has been arisen. Recently, multi-scale analyses have been developed to take into account the primary side as well. In this study, the coupled CUPID and MARS code was used for the simulation of boiler side of the PWR steam generator. Calculation results are compared with the existing code quantitatively. Coupled CUPID/MARS code was applied for the simulation of the steam generator. The primary side of the steam generator and other RCS was simulated by MARS and the secondary side was calculated by CUPID with porous media approach.

Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

International Nuclear Information System (INIS)

Hidaka, Akihide; Asaka, Hideaki; Sugimoto, Jun; Ueno, Shingo; Yoshino, Takehito

1999-12-01

In PWR severe accidents such as station blackout, the integrity of steam generator U-tube would be threatened early at the transient among the pipes of primary system. This is due to the hot leg countercurrent natural circulation (CCNC) flow which delivers the decay heat of the core to the structures of primary system if the core temperature increases after the secondary system depressurization. From a view point of accident mitigation, this steam generator tube rupture (SGTR) is not preferable because it results in the direct release of primary coolant including fission products (FP) to the environment. Recent SCDAP/RELAP5 analyses by USNRC showed that the creep failure of pressurizer surge line which results in release of the coolant into containment would occur earlier than SGTR during the secondary system depressurization. However, the analyses did not consider the decay heat from deposited FP on the steam generator U-tube surface. In order to investigate the effect of decay heat on the steam generator U-tube integrity, the hot leg CCNC flow model used in the USNRC's calculation was, at first, validated through the analysis for JAERI's LSTF experiment. The CCNC model reproduced well the thermohydraulics observed in the LSTF experiment and thus the model is mostly reliable. An analytical study was then performed with SCDAP/RELAP5 for TMLB' sequence of Surry plant with and without secondary system depressurization. The decay heat from deposited FP was calculated by JAERI's FP aerosol behavior analysis code, ART. The ART analysis showed that relatively large amount of FPs may deposit on steam generator U-tube inlet mainly by thermophoresis. The SCDAP/RELAP5 analyses considering the FP decay heat predicted small safety margin for steam generator U-tube integrity during secondary system depressurization. Considering associated uncertainties in the analyses, the potential for SGTR cannot be ignored. Accordingly, this should be considered in the evaluation of merits

Steam Generator Group Project: Task 10, Secondary side examination

International Nuclear Information System (INIS)

Schwenk, E.B.

1987-06-01

This report concludes an effort to examine and assess from the secondary side, the condition of the retired-from-service Surry 2A steam generator. It is includes photographs of degradation of various components or regions in a generic recirculating type steam generator. The photographic detail given in the text (and the previous report NUREG/CR-3843, PNL-5033) have not been readily available to many investigators outside the Nuclear Steam Supply Vendors and Users. The photographs include views of Inconel 600 heat exchanger tubes (0.875 diameter [nominal] x 0.050 inch wall) showing deformed and intergranularly stress-corrosion cracked U-bends, tube denting in the support plate, intergranular attack and thinning (both in the tube sheet region), support plat deformation and cracking at flow slots and in ligaments between flow holes and tube holes. In addition, photographs of tube pitting, anti-vibration bar fretting, and the sludge pile are presented. An experimental stress analysis was conducted on a distorted Row 1 tube in the region of a compressed flow slot between the 6th and 7th (top) support plates. During removal of the tube, relaxation strains were measured and residual stresses calculated. Finally a cursory metallurgical failure analysis was conducted on a broken U-bend (R1C91) to determine its mode of failure. A rotabroach boring technique was used to make multiple penetrations with minimal damage to the secondary side, at various locations on the shell

Steam generating system in LMFBR type reactors

International Nuclear Information System (INIS)

Kurosawa, Katsutoshi.

1984-01-01

Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

Testing installation for a steam generator

International Nuclear Information System (INIS)

Dubourg, M.

1985-01-01

The invention proposes a testing installation for a steam generator associated to a boiler, comprising a testing exchanger connected to a feeding circuit in secondary fluid and to a circuit to release the steam produced, and comprising a heating-tube bundle connected to a closed circuit of circulation of a primary coolant at the same temperature and at the pressure than the primary fluid. The heating-tube bundle of the testing exchanger has the same height than the primary bundle of the steam generator and the testing exchanger is at the same level and near the steam generator and is fed by the same secondary fluid such as it is subject to the same operation phases during a long period. The in - vention applies, more particularly, to the steam generators of pressurized water nuclear power plants [fr

Stress corrosion cracking susceptibility of steam generator tubing on secondary side in restricted flow areas

International Nuclear Information System (INIS)

Fulger, M.; Lucan, D.; Radulescu, M.; Velciu, L.

2003-01-01

Nuclear steam generator tubes operate in high temperature water and on the secondary side in restricted flow areas many nonvolatile impurities accidentally introduced into circuit tend to concentrate. The concentration process leads to the formation of highly aggressive alkaline or acid solutions in crevices, and these solutions can cause stress corrosion cracking (SCC) on stressed tube materials. Even though alloy 800 has shown to be highly resistant to general corrosion in high temperature water, it has been found that the steam generator tubes may crack during service from the primary and/or secondary side. Stress corrosion cracking is still a serious problem occurring on outside tubes in operating steam generators. The purpose of this study was to evaluate the environmental factors affecting the stress corrosion cracking of steam generators tubing. The main test method was the exposure for 1000 hours into static autoclaves of plastically stressed C-rings of Incoloy 800 in caustic solutions (10% NaOH) and acidic chloride solutions because such environments may sometimes form accidentally in crevices on secondary side of tubes. Because the kinetics of corrosion of metals is indicated by anodic polarization curves, in this study, some stressed specimens were anodically polarized in caustic solutions in electrochemical cell, and other in chloride acidic solutions. The results presented as micrographs, potentiokinetic curves, and electrochemical parameters have been compared to establish the SCC behavior of Incoloy 800 in such concentrated environments. (authors)

Secondary side photographic techniques used in characterization of Surry steam generator

International Nuclear Information System (INIS)

Sinclair, R.B.

1984-10-01

Characterization of the generator's secondary side prior to destructive removal of tubing presents a significant challenge. Information must be obtained in a radioactive field (up to 15 R/h) throughout the tightly spaced bundle of steam generator tubes. This report discusses the various techniques employed, along with their respective advantages and disadvantages. The most successful approach to nondestructive secondary side characterization and documentation was through use of in-house developed pinhole cameras. These devices provided accurate photographic documentation of generator condition. They could be fabricated in geometries allowing access to all parts of the generator. Semi-remote operation coupled with large area coverage per investigation and short at-location times resulted in significant personnel exposure advantages. The fabrication and use of pinhole cameras for remote inspection is discussed in detail

Predicted wear on the tube outside surface due to foreign object in the secondary side of steam generator

International Nuclear Information System (INIS)

Kim, Hyung Nam; Cho, Nam Cheoul

2012-01-01

It is necessary to evaluate the effects of foreign objects on steam generator tubes and to use this information to take appropriate safety precautions to prevent nuclear accidents. Foreign objects may include loose parts from the feed water system and items lost by workers during o/h, and may flow into the secondary side of steam generators during operation. A foreign object could damage steam generator tube walls if there is relative motion between the tube and the foreign object. This is especially true for foreign objects that land on the tube sheet because the velocity of cross flow, which creates a contact force between the tube and foreign object, is relatively high there. During steam generator overhauls, foreign objects are detected by non destructive methods such as the visual test and/or the eddy current test. Confirmed foreign objects should be removed for nuclear safety. The Foreign Object Search and Retrieval System (FOSAR) can be used to remove foreign objects from the steam generators with a square tube array. However, the FOSAR cannot be used (or can be used in only a very restricted area, such as the outside of the tube bundle) in the steam generators with a triangular tube array. In order to continue nuclear power plant operations without removing foreign objects, the integrity of the steam generator tube must be verified. This paper introduces a practical method developed to evaluate the effects of foreign objects detected on tube sheets in the secondary sides of steam generators

Carbon steel corrosion prevention during chemical cleaning of steam generator secondary side components

International Nuclear Information System (INIS)

Fulger, M.; Lucan, D.; Velciu, L.

2009-01-01

During operation of a nuclear power plant, many contaminants, such as solid particles or dissolved species are formed in the secondary circuit, go into steam generator and deposit as scales on heat transfer tubing, support plate or as sludge on tube sheet. By accumulation of these impurities, heat transfer is reduced and the integrity of the steam generator tubing is influenced. Chemical cleaning is a qualified, efficient measure to improve steam generator corrosion performance. The corrosion mechanism can be counteracted by the chemical cleaning of the deposits on the tube sheet and the scales on the heat transfer tubing. The major component of the scales is magnetite, which can be dissolved using an organic chelating agent (ethylenediaminetetraacetic acid, EDTA) in combination with a complexing agent such as citric acid in an alkaline reducing environment. As the secondary side of SG is a conglomerate of alloys it is necessary to choose an optimal chemical cleaning solution for an efficient cleaning properties and at the same time with capability of corrosion prevention of carbon steel components during the process. The paper presents laboratory tests initiated to confirm the ability of this process to clean the SG components. The experiments followed two paths: - first, carbon steel samples have been autoclavized in specific secondary circuit solutions of steam generator to simulate the deposits constituted during operation of this equipment; - secondly, autoclavized samples have been cleaned with a solvent composed of EDTA citric acid, hydrazine of pH = 5 and temperature of 85 deg. C. Before chemical cleaning, the oxide films were characterized by surface analysis techniques including optical microscopy, scanning electron microscopy (SEM) and electrochemical impedance spectroscopy (EIS). Applied to dissolve corrosion products formed in a steam generator, the solvents based on chelating agents are aggressive toward carbon steels and corrosion inhibitors are

A study on the Stress Corrosion Cracking reduction method of Steam Generator secondary side of KSNP

International Nuclear Information System (INIS)

Kim, June Hoon; Lee, Goune Jin

2014-01-01

In order to avoid sludge accumulation affecting the life of the steam generator, the best way is to prevent the sludge inflow in advance by optimization of water quality management through chemical concentration and pH control etc. However it is very difficult to prevent sludge accumulation under the weak condition of corrosion, such as condensation, boiling and high temperature of feed-water in NPPs. Particularly stress corrosion cracking occurs in a top-of-tube sheet area of steam generator with an increase in number of operation years of Korea Standard Nuclear Plant(KSNP)... The purpose of this study is to improve suppression of stress corrosion cracking and life extension for steam generator and improve plant efficiency by performing full length bulk high chemical cleaning in order to remove iron oxide of steam generator secondary side in KSNP Hanbit Unit 6. This study analyzed the Free EDTA and Fe concentrations and sludge removal after performed full length bulk high temperature chemical cleaning for removing the iron oxide of steam generator secondary side, which of Hanbit unit 6 of KSNP. 1) It showed a typical pattern that Fe concentration increased in accordance with to decrease Free EDTA(Ethylene Diamine Tetea acetic Acid) concentration. 2) Sludge removal based on iron oxide after performing the full length bulk high temperature chemical cleaning was 3001kg and sludge removal by lancing additionally was 200.1kg

Three-dimensional modeling of nuclear steam generator

International Nuclear Information System (INIS)

Bogdan, Z.; Afgan, N.

1985-01-01

In this paper mathematical model for steady-state simulation of thermodynamic and hydraulic behaviour of U-tube nuclear steam generator is described. The model predicts three-dimensional distribution of temperatures, pressures, steam qualities and velocities in the steam generator secondary loop. In this analysis homogeneous two phase flow model is utilized. Foe purpose of the computer implementation of the mathematical model, a special flow distribution code NUGEN was developed. Calculations are performed with the input data and geometrical characteristics related to the D-4 (westinghouse) model of U-tube nuclear steam generator built in Krsko, operating under 100% load conditions. Results are shown in diagrams giving spatial distribution of pertinent variables in the secondary loop. (author)

Characterization and dissolution studies of Bruce Unit 3 steam generator secondary side deposits

International Nuclear Information System (INIS)

Semmler, J.

1998-01-01

The physical and chemical properties of secondary side steam generator deposits in the form of powder and flake obtained from Bruce Nuclear Generating Station A (BNGS A) Unit 3 were studied. The chemical phases present in both types of deposits, collected prior to the 1994 chemical cleaning during the pre-clean water lancing campaign, were magnetite (Fe 3 O 4 ), metallic copper (Cu), hematite (Fe 2 O 3 ) and cuprous oxide (Cu 2 O). The major difference between the chemical composition of the powder and the flake was the presence of zinc silicate (Zn 2 SiO 4 ) and several unidentified silicate phases containing Ca, Al, Mn, and Mg in the flake. The flake deposit had high hardness values, high electrical resistivity, low porosity and a lower dissolution rate in the EPRI-SGOG (Electric Power Research Institute-Steam Generator Owner's Group) chemical cleaning solvents compared to the powder deposit. Differences in the deposit properties after chemical cleaning of the Unit 3 steam generators and after laboratory cleaning were noted. The presence of silicates in the deposit inhibit magnetite dissolution

Aerosol trapping in steam generator (artist): an investigation of aerosol and iodine behaviour in the secondary side of a steam generator

International Nuclear Information System (INIS)

Guentay, S.; Birchley, J.; Suckow, D.; Dehbi, A.

2000-01-01

Incidents such as a steam generator tube rupture (SGTR) with stuck-open relief valve are important accident sequences for analysis by virtue of the open path for release of radioactivity which ensues. The release may be mitigated by deposition of fission products on the steam generator (SG) tubes and other structures, or by scrubbing in the secondary coolant. The absence of empirical data, the complexity of the geometry and controlling processes, however, make the retention difficult to quantify and its full import is typically not taken into account in risk assessment studies. The ARTIST experimental programme at PSI will simulate the flow and retention of aerosol-borne fission products in the SG secondary, and thus provide a unique database to support safety assessments and analytical models. Scaling of the break flow represents a particular challenge since the aerosol retention processes operate at contrasting length scales. Preliminary calculations have identified a baseline set of conditions, and confirmed the feasibility of the rig design and scaling principles. Flexibility of the rig layout enables simulations to be performed for a range of SG designs, accident situations and accident management philosophies. (authors)

Field experience of cleanliness inspection for secondary-side in PWR steam generators

International Nuclear Information System (INIS)

Ding Xunshen

1997-05-01

The mechanical cleaning and TV inspection technology for secondary-side in steam generators of Daya Bay Nuclear Power Plant has been used for preventing the heat transfer tubes from damage caused by residues in steam generator and foreign objects. A lancing has been used for steam generators. The high-pressure jet sent from the central lane to the inside of the bundle has two objectives: (a) the mechanical energy in the jet breaks up the deposit, and (b) the particles are then carried to the periphery of the generator, where they are collected by a water circulation system. The TV inspection consists of the inspection before lancing and after lancing. The former includes the inspection of outer tube lane and central tube lane, and the extraction of foreign objects; the latter includes the inspection of inter tube area, outer tube lane and central tube lane, and the extraction of foreign objects. Video cassette are visualized by a specialist who is qualified to realize a map representative of the cleanliness state of the tubesheet, and to judge if it is or not acceptable. The TV inspection obtained a cleanliness conclusion according to acceptance criteria. So, it is important to work out cleanliness acceptance criteria suit for every operation stages. The site practices shown that after lancing, although hard deposit existed and height of hard deposit in some local place exceeded 5 mm, but the sludge remaining on tubesheet was less. If we can conduct periodic lancing to steam generator, the increase rate of hard deposit will be small. (7 refs., 12 figs.)

Horizontal steam generator PGV-1000 thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Ubra, O. [Skoda Company, Prague (Switzerland); Doubek, M. [Czech Technical Univ., Prague (Switzerland)

1995-12-31

A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.

Horizontal steam generator PGV-1000 thermal-hydraulic analysis

International Nuclear Information System (INIS)

Ubra, O.; Doubek, M.

1995-01-01

A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.)

Horizontal steam generator PGV-1000 thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Ubra, O [Skoda Company, Prague (Switzerland); Doubek, M [Czech Technical Univ., Prague (Switzerland)

1996-12-31

A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.

Development of Program Evaluating the Effects on the Secondary Side of Steam Generator due to Foreign Objects

International Nuclear Information System (INIS)

Ju, Yoo Hyun; Nam, Choi Sung

2005-01-01

When materials such as metal are into the secondary side of steam generator, they, so called foreign objects, may have influences on the integrity of the steam generator tubes. They cause the tube wear due to the relative motion between the tubes and foreign objects and the tube impact due to flow. The best way to avoid the effects is to remove all the foreign objects. However, it is not easy to remove the foreign materials thoroughly due to their condition such as the location. If the locations of the foreign materials are in the middle of tube bundle and the tube arrangement of the steam generator is the triangle type, the equipment such as FOSAR(Foreign Object Search and Retrieval) can not reach their locations. If the foreign materials stick together with the tubes or tube sheet, they can not be removed. In the case of operating the steam generator with the foreign materials, the licensee must prove that the materials do not affect the tube integrity and do not threaten the pressure boundary with the analytical method. Considering the wear and impact by the foreign materials, KEPRI(Korea Electric Power Research Institute) developed the methodology to evaluate the foreign materials analytically. This methodology was described with a computer program in order to obtain the fast results. The program informs whether the tubes have the structural integrity when the foreign material strikes the tubes. Moreover, this gives us the remaining life of the steam generator tubes. In this paper, the program, which evaluates the effects of the foreign objects in the secondary side of steam generator, is introduced

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

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

Corrosion Evaluation and Corrosion Control of Steam Generators

International Nuclear Information System (INIS)

Maeng, W. Y.; Kim, U. C.; Sung, K. W.; Na, J. W.; Lee, Y. H.; Lee, D. H.; Kim, K. M.

2008-06-01

Corrosion damage significantly influences the integrity and efficiency of steam generator. Corrosion problems of steam generator are unsolved issues until now even though much effort is made around world. Especially the stress corrosion cracking of heat exchange materials is the first issue to be solved. The corrosion protection method of steam generator is important and urgent for the guarantee of nuclear plant's integrity. The objectives of this study are 1) to evaluate the corrosion properties of steam generator materials, 2) to optimize the water chemistry of steam generator and 3) to develop the corrosion protection method of primary and secondary sides of steam generator. The results will be reflected to the water chemistry guideline for improving the integrity and efficiency of steam generator in domestic power plants

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

Secondary coolant circuit operation tests: steam generator feedwater supply

International Nuclear Information System (INIS)

Beroux, M.

1985-01-01

No one important accident occurred during the start-up tests of the 1300MWe P4 series, concerning the feedwater system of steam generators (SG). This communication comments on some incidents, that the tests allowed to detect very soon and which had no consequences on the operation of units: 1) Water hammer in feedwater tubes, and incidents met in the emergency steam generator water supply circuit. The technological differences between SG 900 and 1300 are pointed out, and the measures taken to prevent this problem are presented. 2) Incidents met on the emergency feedwater supply circuit of steam generators; mechanical or functional modifications involved by these incidents [fr

Steam generator operating experience update, 1982-1983

International Nuclear Information System (INIS)

Frank, L.

1984-06-01

This report is a continuation of earlier reports by the staff addressing pressurized water reactor steam generator operating experience. NUREG-0886, Steam Generator Tube Experience, published in February 1982 summarized experience in domestic and foreign plants through December 1981. This report summarizes steam generator operating experience in domestic plants for the years 1982 and 1983. Included are new problems encountered with secondary-side loose parts, sulfur-induced stress-assisted corrosion cracking, and flow-induced vibrational wear in the new preheater design steam generators. The status of Unresolved Safety Issues A3, A4, and A5 is also discussed

Corrosion Evaluation and Corrosion Control of Steam Generators

Energy Technology Data Exchange (ETDEWEB)

Maeng, W. Y.; Kim, U. C.; Sung, K. W.; Na, J. W.; Lee, Y. H.; Lee, D. H.; Kim, K. M

2008-06-15

Corrosion damage significantly influences the integrity and efficiency of steam generator. Corrosion problems of steam generator are unsolved issues until now even though much effort is made around world. Especially the stress corrosion cracking of heat exchange materials is the first issue to be solved. The corrosion protection method of steam generator is important and urgent for the guarantee of nuclear plant's integrity. The objectives of this study are 1) to evaluate the corrosion properties of steam generator materials, 2) to optimize the water chemistry of steam generator and 3) to develop the corrosion protection method of primary and secondary sides of steam generator. The results will be reflected to the water chemistry guideline for improving the integrity and efficiency of steam generator in domestic power plants.

Wear on Plugged Tube due to the Foreign Objects on the Secondary Side of Steam Generator

International Nuclear Information System (INIS)

Kim, Hyung Nam; Cho, Nam Cheoul; Nam, Min Woo

2013-01-01

In this paper, the changes of the tube frequency and amplitude are introduced before and after plugging. The amplitude of the bottom span for the steam generator tube is not much changed after tube plugging. Moreover, the contact force between the plugged tube and the foreign object is the same as that of intact tube and the foreign object. However, the frequencies of plugged tubes are about 9∼12% higher than those of intact tubes. That means the wear due to the foreign object would be accelerated after the tube plugging. Therefore, the tube stabilizer should be installed when the tube is plugged due to the foreign object wear. The tube wall of steam generator is a pressure boundary between the coolant of the primary system and the feedwater of the secondary system. It is very important to insure the structural integrity of the tubes because the radioactive coolant is flow into the feedwater due to the pressure difference as the result of tube failure. The degradations of steam generator tubes are corrosion, wear, fatigue and foreign object wear, etc. The foreign object wear is one of mechanical degradation due to materials flew into the secondary side of steam generator. The steam generator tubes, estimated not to insure structural integrity from the results of the nondestructive evaluation such as eddy current test and visual inspection, are excluded from the service with plugging. However, the tube wear is still being progressed after the plugging because the relative motion between the tube and structure is still existed due to the secondary side flow in the steam generator. If the tube is completely cut because of the degradation, the tube can be a stress or of failure of tubes around the plugged tube. The contact force between the structure and tube is lowered as the wear is progressed. However, the contact force between the foreign object and tube is not changed as the wear is progressed. Therefore, the structural integrity of tubes around the foreign

Surry Power Station secondary water chemistry improvement since steam generator replacement and the unit two feedwater pipe rupture

International Nuclear Information System (INIS)

Swindell, E.T.

1988-01-01

Surry Power Station has two Westinghouse-designed three-loop PWRs of 811 MWe design rating. The start of commercial operation was in July, 1972 in No.1 plant, and May, 1973 in No.2 plant. Both plants began the operation using controlled phosphate chemistry for the steam generators. In 1975, both plants were converted to all volatile treatment on the secondary side due to the tube wall thinning corrosion in the steam generators, which was associated with the phosphate sludge that was building up on the tube sheets and created acidic condition. Thereafter, condenser and air leakage and steam generator denting occurred, and after the operation of 8 years 2 month of No.1 plant and 5 years 9 months of No.2 plant, the steam generators were replaced. A major plant improvement program was designed and implemented from 1979 to 1980. The improvement in new steam generators, the modification for preventing corrosion, the addition of a steam generator blowdown recovery system, the reconstruction of condensers, the installation of full flow, deep bed condensate polishers, the addition of Dionex 8,000 on-line ion chromatograph, a long term maintenance agreement with Westinghouse and so on are reported. (Kako, I.)

Third international seminar on horizontal steam generators

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-12-31

The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues.

Third international seminar on horizontal steam generators

International Nuclear Information System (INIS)

1995-01-01

The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues

Third international seminar on horizontal steam generators

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-12-31

The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues.

Predictive models for deposit accumulation and corrosion on secondary side of steam generators

International Nuclear Information System (INIS)

Choi, Samuel; Moroney, Velvet; Marks, Chuck; Kreider, Marc

2012-09-01

Experience demonstrates that deposit accumulation in steam generators (SGs) can lead to corrosion of tubes. To minimize the probability of this corrosion, utilities employ a variety of deposit control strategies. However, these processes can involve significant costs and potentially affect outage critical paths. Since there has been no model that quantifies tube degradation as a function of deposit accumulation, utilities have had to make decisions regarding deposit control strategies without a reliable quantitative basis. The objective of this study is to develop methods that utilities can use to quantify benefits of SG deposit control strategies with regard to rates of secondary-side tube corrosion. Two different methodologies are employed in this work. The first methodology is empirical and is involved an attempt to correlate degradation rates with deposit accumulation as indicated by sludge pile height. Because there has been relatively little tube degradation in currently operating steam generators, this correlation is developed using data for Alloy 600MA SG tubes. To increase the number of units that could be used for defect/deposit correlations, a method to relate the sludge pile deposit mass and the number of tubes with non-zero sludge height is developed. The second methodology is theoretical and is based on the use of calculated differences in temperature and chemistry to predict the effect of deposits on corrosion rates. Computational fluid dynamics (CFD) models are developed to simulate thermal-hydraulic conditions representative of conditions that are present within porous deposits formed at the top of tube sheet. This paper will discuss the development and application of the predictive models for deposit accumulation and corrosion on the secondary side of steam generators. (authors)

Steam generator operation and maintenance

International Nuclear Information System (INIS)

Lee, C.K.

2004-01-01

capacity due to excessive pressure drop across the tube support plates. OTSG owners group has developed both mechanical and chemical cleaning process and an upgraded secondary water chemistry in resolving these problems. The OTSG performance has been greatly improved since OTSG plants implemented chemical cleaning and morpholine water chemistry. The SGOG project officially ended December 31, 1986. A six-year Steam Generator Reliability Program (SGRP) under the EPRI base program began January 1, 1987. SGRP continued to address the generic steam generator problems facing nuclear utilities. In order to develop appropriate strategies to cope with the tube degradation problems, SGRP has performed the statistical evaluations to model the progression of damage mechanism aimed at accurate prediction of the defect growth rate of various mechanisms such that long term trends can be developed. Analysis of the behavior of group of plants indicate that insights on the potential behavior of a specific plant may be developed from the observed behavior at other plants. SGRP has provided utilities with tube inservice inspection guidelines (ISI Guideline) including ISI Performance Demonstration program to help utilities to improve tube inspection accuracy and sensitivity. SGRP has also updated the secondary Chemistry Guidelines and worked on the advanced amine application guidelines to better protect the steam generator tube corrosion

Modeling and Simulation of U-tube Steam Generator

Science.gov (United States)

Zhang, Mingming; Fu, Zhongguang; Li, Jinyao; Wang, Mingfei

2018-03-01

The U-tube natural circulation steam generator was mainly researched with modeling and simulation in this article. The research is based on simuworks system simulation software platform. By analyzing the structural characteristics and the operating principle of U-tube steam generator, there are 14 control volumes in the model, including primary side, secondary side, down channel and steam plenum, etc. The model depends completely on conservation laws, and it is applied to make some simulation tests. The results show that the model is capable of simulating properly the dynamic response of U-tube steam generator.

Operating experience with steam generator water chemistry in Japanese PWR plants

International Nuclear Information System (INIS)

Onimura, K.; Hattori, T.

1991-01-01

Since the first PWR plant in Japan started its commercial operation in 1970, seventeen plants are operating as of the end of 1990. First three units initially applied phosphate treatment as secondary water chemistry control and then changed to all volatile treatment (AVT) due to phosphate induced wastage of steam generator tubing. The other fourteen units operate exclusively under AVT. In Japan, several corrosion phenomena of steam generator tubing, resulted from secondary water chemistry, have been experienced, but occurrence of those phenomena has decreased by means of improvement on impurity management, boric acid treatment and high hydrazine operation. Recently secondary water chemistry in Japanese plants are well maintained in every stage of operation. This paper introduces brief summary of the present status of steam generators and secondary water chemistry in Japan and ongoing activities of investigation for future improvement of reliability of steam generator. History and present status of secondary water chemistry in Japanese PWRs were introduced. In order to get improved water chemistry, the integrity of secondary system equipments is essential and the improvement in water chemistry has been achieved with the improvement in equipments and their usage. As a result of those efforts, present status of secondary water is excellent. However, further development for crevice chemistry monitoring technique and an advanced water chemistry data management system is desired for the purpose of future improvement of reliability of steam generator

An expert system for steam generator maintenance

International Nuclear Information System (INIS)

Remond, A.

1988-01-01

The tube bundles in PWR steam generators are, by far, the major source of problems whether they are due to primary and secondary side corrosion mechanisms or to tube vibration-induced wear at tube support locations. Because of differences in SG operating, materials, and fabrication processes, the damage may differ from steam generator to steam generator. MPGV, an expert system for steam generator maintenance uses all steam generator data containing data on materials, fabrication processes, inservice inspection, and water chemistry. It has access to operational data for individual steam generators and contains models of possible degradation mechanisms. The objectives of the system are: · Diagnosing the most probable degradation mechanism or mechanisms by reviewing the entire steam generator history. · Identifying the tubes most exposed to future damage and evaluating the urgency of repair by simulating the probable development of the problem in time. · Establishing the appropriate preventive actions such as repair, inspection or other measures and establishing an action schedule. The system is intended for utilities either for individual plants before each inspection outage or any time an incident occurs or for a set of plants through a central MPGV center. (author)

PMK-2. Experimental study on steam generator behaviour

Energy Technology Data Exchange (ETDEWEB)

Ezsoel, G.; Szabados, L.; Trosztel, I. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

1995-12-31

The PMK-2 is a full pressure scaled-down model of the Paks Nuclear Power Plant, with a 1:2070 scaling ratio for the volume and power. It has a steam generator model which is a vertical section of the horizontal steam generator. The model has hot and cold collectors similarly to the steam generators of the plant. The heat transfer tubes are horizontal tubes. There are 82 rows of tubes and the elevations, as well as the heat transfer surface distribution is the same as in the plant. The elevation of the feed water supply is similar to that of the plant. To study the temperature distribution in both the primary and the secondary side several thermocouples are built in, in addition to the overall instrumentation of the loop which has again a high number of measurement channels. Paper gives a description and results of SPE-4, with special respect to the steam generator behaviour in both steady state and transient conditions. Axial distribution of coolant and feedwater temperatures are given for the primary and the secondary side of hot and cold collectors and the temperature distribution in the centre of steam generator. (orig.).

PMK-2. Experimental study on steam generator behaviour

Energy Technology Data Exchange (ETDEWEB)

Ezsoel, G; Szabados, L; Trosztel, I [KFKI Atomic Energy Research Inst., Budabest (Hungary)

1996-12-31

The PMK-2 is a full pressure scaled-down model of the Paks Nuclear Power Plant, with a 1:2070 scaling ratio for the volume and power. It has a steam generator model which is a vertical section of the horizontal steam generator. The model has hot and cold collectors similarly to the steam generators of the plant. The heat transfer tubes are horizontal tubes. There are 82 rows of tubes and the elevations, as well as the heat transfer surface distribution is the same as in the plant. The elevation of the feed water supply is similar to that of the plant. To study the temperature distribution in both the primary and the secondary side several thermocouples are built in, in addition to the overall instrumentation of the loop which has again a high number of measurement channels. Paper gives a description and results of SPE-4, with special respect to the steam generator behaviour in both steady state and transient conditions. Axial distribution of coolant and feedwater temperatures are given for the primary and the secondary side of hot and cold collectors and the temperature distribution in the centre of steam generator. (orig.).

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.)

Proceedings of steam generator sludge deposition in recirculating and once through steam generator upper tube bundle and support plates

International Nuclear Information System (INIS)

Baker, R.L.; Harvego, E.A.

1992-01-01

The development of remedial measures of shot peening have given nuclear utilities viable measures to address primary water stress corrosion cracking to extend steam generator life. The nuclear utility industry is now faced with potential replacement of steam generators in nuclear power plants due to stress corrosion cracking and intergranular attach in crevice locations on the secondary side of steam generators at tube support plates and at the crevice at the top of the tube sheet. Significant work has been done on developing and understanding of the effects of sludge buildup on the corrosion process at these locations. This session was envisioned to provide a forum for the development of an understanding of the mechanisms which control the transport and deposition of sludge on the secondary side of steam generators. It is hoped that this information will aid utilities in monitoring the progression of fouling of these crevices by further knowledge in where to look for the onset of support plate crevice fouling. An understanding of the progression of fouling from upper tube support plates to those lower in the steam generator where higher temperatures cause the corrosion process to initiate first can aid the nuclear utility industry in developing remedial measures for this condition and in providing a forewarning of when to apply such remedial measures

Steam generators secondary side chemical cleaning at Point Lepreau using the Siemen's high temperature process

International Nuclear Information System (INIS)

Verma, K.; MacNeil, C.; Odar, S.

1996-01-01

The secondary sides of all four steam generators at the Point Lepreau Nuclear Generating Stations were cleaned during the 1995 annual outage run-down using the Siemens high temperature chemical cleaning process. Traditionally all secondary side chemical cleaning exercises in CANDU as well as the other nuclear power stations in North America have been conducted using a process developed in conjunction with the Electric Power Research Institute (EPRI). The Siemens high temperature process was applied for the first time in North America at the Point Lepreau Nuclear Generating Station (PLGS). The paper discusses experiences related to the pre and post award chemical cleaning activities, chemical cleaning application, post cleaning inspection results and waste handling activities. (author)

Digital simulation for nuclear once-through steam generators

International Nuclear Information System (INIS)

Chen, A.T.

1976-01-01

Mathematical models for calculating the dynamic response of the Oconee type once through steam generator (OTSG) and the integral economizer once through steam generator (IEOTSG) was developed and presented in this dissertation. Linear and nonlinear models of both steam generator types were formulated using the state variable, lumped parameter approach. Transient and frequency responses of system parameters were calculated for various perturbations from both the primary coolant side and the secondary side. Transients of key parameters, including primary outlet temperature, superheated steam outlet temperature, boiling length/subcooled length and steam pressure, were generated, compared and discussed for both steam generator types. Frequency responses of delta P/sub s//deltaT/sub pin/ of the linear OTSG model were validated by using the dynamic testing results obtained at the Oconee I nuclear power station. A sensitivity analysis in both the time and the frequency domains was performed. It was concluded that the mathematical and computer models developed in this dissertation for both the OTSG and the IEOTSG are suitable for overall plant performance evaluation and steam generator related component/system design analysis for nuclear plants using either type of steam generator

Strategy for maintaining cleanliness in the secondary part of Steam Generators of French PWRs fleet

International Nuclear Information System (INIS)

Prin, C.

2012-01-01

Between 2004 and 2006, EDF noticed primary / secondary leakages on Steam Generators (SG) of a French nuclear plant. Further studies have shown that fouling by iron oxides and clogging of the quatrefoils broached holes of the Tubes Support Plates (TSP) are aggravating factors of the risk of cracking by vibration-induced fatigue of the SG's tubes. EDF then initiated a program of chemical cleaning of steam generators to restore and maintain an acceptable state of cleanliness of the secondary part of SG. This paper presents EDF's strategy to achieve this goal. It is organized as follows : first, fouling and clogging phenomenon and associated estimation means are explained. Then the strategy of chemical cleaning and its evolution are presented. The implemented processes as well as process monitoring and plants requalification are also described. Finally, the operations performed, the continuous improvement process conducted in connection with the French Nuclear Safety Authority (ASN) and the control of environmental impact are reviewed. This paper was written with the support of AREVA and Westinghouse, the two operators of chemical cleaning in France. (author)

Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

International Nuclear Information System (INIS)

Pappx, L.

1994-01-01

After modification of Dukovany NPP steam generator feedwater system, the increased concentration of minerals was measured in the cold leg of modified steam generator. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators, has focused this attention on the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of flow distribution in the secondary side of SG was developed. (Author)

Resistance of Incoloy 800 steam generator tube to pitting corrosion in PWR secondary water at 250°C

International Nuclear Information System (INIS)

Schvartzman, Mônica M.A.M.; Albuquerque, Adriana Silva de; Esteves, Luiza; Rabello, Emerson G.; Mansur, Fábio Abud

2017-01-01

The steam generator (SG) is one of the main components of a PWR, so the performance of this type of nuclear power plant depends to a large extent on the trouble-free operation of SGs. Its degradation significantly affects the overall plant performance. Alloy 800NG (Incoloy® 800) is a nickel-iron-chromium alloy used for steam generator tubes in PWRs due to their high strength, good workability and resistance to corrosion. This behavior is attributed to the protective oxide film formed on the metal surface by contact with the high temperature pressurized water. However, chloride is one of major SG impurities that cause the breakdown of the passive film and initiate localized corrosion in passive metals as Alloy 800NG. The aim of this study is to provide information about the pitting corrosion behavior of the Incoloy® 800 steam generator tube under normal secondary circuit parameters (250 deg C and 5 MPa) and abnormal conditions of operation (presence of chloride ions in the secondary water). For this, optical microscopy, XRD and EDS analysis and electrochemical tests have been carried out under simulated PWR secondary water operating conditions. The susceptibility to pitting corrosion was evaluated using electrochemical tests and the oxide layer formed on material was examined by means of scanning electron microscopy (SEM) and energy dispersive X-ray spectrometer (EDS) analyses. (author)

Resistance of Incoloy 800 steam generator tube to pitting corrosion in PWR secondary water at 250°C

Energy Technology Data Exchange (ETDEWEB)

Schvartzman, Mônica M.A.M. [Pontifícia Universidade Católica de Minas Gerais (PUC-Minas), Belo Horizonte, MG (Brazil); Albuquerque, Adriana Silva de; Esteves, Luiza; Rabello, Emerson G.; Mansur, Fábio Abud, E-mail: monicacdtn@gmail.com, E-mail: asa@cdtn.br, E-mail: luiza.esteves@cdtn.br, E-mail: egr@cdtn.br, E-mail: fametalurgica@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

The steam generator (SG) is one of the main components of a PWR, so the performance of this type of nuclear power plant depends to a large extent on the trouble-free operation of SGs. Its degradation significantly affects the overall plant performance. Alloy 800NG (Incoloy® 800) is a nickel-iron-chromium alloy used for steam generator tubes in PWRs due to their high strength, good workability and resistance to corrosion. This behavior is attributed to the protective oxide film formed on the metal surface by contact with the high temperature pressurized water. However, chloride is one of major SG impurities that cause the breakdown of the passive film and initiate localized corrosion in passive metals as Alloy 800NG. The aim of this study is to provide information about the pitting corrosion behavior of the Incoloy® 800 steam generator tube under normal secondary circuit parameters (250 deg C and 5 MPa) and abnormal conditions of operation (presence of chloride ions in the secondary water). For this, optical microscopy, XRD and EDS analysis and electrochemical tests have been carried out under simulated PWR secondary water operating conditions. The susceptibility to pitting corrosion was evaluated using electrochemical tests and the oxide layer formed on material was examined by means of scanning electron microscopy (SEM) and energy dispersive X-ray spectrometer (EDS) analyses. (author)

Planning of the steam generators for nuclear applications using optimization techniques

International Nuclear Information System (INIS)

Sakai, M.; Silvares, O.M.

1978-01-01

Procedure for the maximization of the net power of a nuclear power plant through the application of the optimal control theory of dynamic systems is presented. The problem is formulated in the steam generator which links the primary and the secondary cycle. The solution of the steam generator, optimization problem is obtained simultaneously with the heat balance in both primary and secondary cycle, through an iterative process. By this way the optimal parameters are obtained for the steam generator, the vapor and the cooling gas cycle [pt

Automatic system for redistributing feedwater in a steam generator of a nuclear power plant

International Nuclear Information System (INIS)

Fuoto, J.S.; Crotzer, M.E.; Lang, G.E.

1980-01-01

A system is described for automatically redistributing a steam generator secondary tube system after a burst in the secondary tubing. This applies to a given steam generator in a system having several steam generators partially sharing a common tube system, and employs a pressure control generating an electrical signal which is compared with given values [fr

Cleanliness criteria to improve steam generator performance

International Nuclear Information System (INIS)

Schwarz, T.; Bouecke, R.; Odar, S.

2005-01-01

High steam generator performance is a prerequisite for high plant availability and possible life time extension. The major opponent to that is corrosion and fouling of the heating tubes. Such steam generator degradation problems arise from the continuous ingress of non-volatile contaminants, i.e. corrosion products and salt impurities may accumulate in the steam generators. These impurities have their origin in the secondary side systems. The corrosion products generally accumulate in the steam generators and form deposits not only in the flow restricted areas, such as on top of tube sheet and tube support structure, but also build scales on the steam generator heating tubes. In addition, the tube scales in general affect the steam generator thermal performance, which ultimately causes a reduction of power output. The most effective ways of counteracting all these degradation problems, and thus of improving the steam generator performance is to keep them in clean conditions or, if judged necessary, to plan cleaning measures such as mechanical tube sheet lancing or chemical cleaning. This paper presents a methodology how to assess the cleanliness condition of a steam generator by bringing together all available operational and inspection data such as thermal performance and water chemistry data. By means of this all-inclusive approach the cleanliness condition is quantified in terms of a fouling index. The fouling index allows to monitor the condition of a specific steam generator, compare it to other plants and, finally, to serve as criterion for cleaning measures such as chemical cleaning. The application of the cleanliness criteria and the achieved field results with respect to improvements of steam generator performance will be presented. (author)

Heat transfer characteristics of horizontal steam generators under natural circulation conditions

International Nuclear Information System (INIS)

Hyvaerinen, J.

1996-01-01

This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent limitation of the decay energy removal in VVER steam generators. The limitation arises as a consequence of the steam generator tube bundle geometry. As an example, it is shown that the grace period associated with pressurizer safety valve opening during a station black-out is 2 1/2-3 hours instead of the 5-6 hours reported in several earlier studies. (However, the change in core heat-up timing is much less-about 1 h at most.) The heat transfer limitation explains the fact that, in the Greifswald VVER 440 station black-out accident in 1975, the steam generators never boiled dry. In addition, the stability of single-phase natural circulation is discussed and insights on the modelling of horizontal steam generators with general-purpose thermal-hydraulic system codes are also presented. (orig.)

Thermal-hydraulic characteristic of the PGV-1000 steam generator

International Nuclear Information System (INIS)

Ubra, O.; Doubek, M.

1995-01-01

Horizontal steam generators are typical parts of nuclear power plants with pressure water reactor type VVER. By means of this computer program, a detailed thermal-hydraulic study of the horizontal steam generator PGV-1000 has been carried out and a special attention has been paid to the thermal-hydraulics of the secondary side. A set of important steam generator characteristics has been obtained and analyzed. Some of the interesting results of the analysis are presented in the paper. (author)

Status of steam generator tubing integrity at Jaslovske Bohunice NPP

International Nuclear Information System (INIS)

Cepcek, S.

1997-01-01

Steam generator represents one of the most important component of nuclear power plants. Especially, loss of tubing integrity of steam generators can lead to the primary coolant leak to secondary circuit and in worse cases to the unit shut down or to the PTS events occurrence. Therefore, to ensure the steam generator tubing integrity and the current knowledge about tube degradation propagation and development is of the highest importance. In this paper the present status of steam generator tubing integrity in operated NPP in Slovak Republic is presented

Three Mile Island Nuclear Station steam generator chemical cleaning

International Nuclear Information System (INIS)

Hansen, C.A.

1992-01-01

The Three Mile Island-1 steam generators were chemically cleaned in 1991 by the B and W Nuclear Service Co. (BWNS). This secondary side cleaning was accomplished through application of the EPRI/SGOG (Electric Power Research Institute - Steam Generator Owners Group) chemical cleaning iron removal process, followed by sludge lancing. BWNS also performed on-line corrosion monitoring. Corrosion of key steam generator materials was low, and well within established limits. Liquid waste, subsequently processed by BWNS was less than expected. 7 tabs

Dynamic simulation of steam generator failures

Energy Technology Data Exchange (ETDEWEB)

Meister, G [Institut fuer Nukleare Sicherheitsforschung, Kernforschungsanlage Juelich GmbH, Juelich (Germany)

1988-07-01

A computer program will be described which is capable to simulate severe transients in a gas heated steam generator. Such transients may arise in the safety analysis of accidents resulting from failures in the heat removal system of an HTGR power plant. Important failure modes which have to be considered are ruptures of one or more steam generator tubes leading to water or steam ejection into the primary system or anomalous operating conditions which my cause damage due to excessive thermal stress. Examples are the complete dryout as a consequence of feedwater interrupt in connection with continuing gas heating and the reflooding of the secondary channel with cold feedwater after dryout. The steam generator program which is capable to simulate accidents of this type is written as a module which can be implemented into a program system fur the simulation of the total heat rejection system. It based on an advanced mathematical model for the two phase flow taking deviations from thermal equilibrium into account. Mass, energy and momentum balances for the primary and secondary fluid and the heat diffusion equations for the heat exchanging wall form a system of coupled differential equations which is solved numerically by an algorithm which is stiffly stable and suppresses effectively oscillations of numerical origin. Results of the simulation of transients of the type mentioned above will be presented and discussed. (author)

Dynamic simulation of steam generator failures

International Nuclear Information System (INIS)

Meister, G.

1988-01-01

A computer program will be described which is capable to simulate severe transients in a gas heated steam generator. Such transients may arise in the safety analysis of accidents resulting from failures in the heat removal system of an HTGR power plant. Important failure modes which have to be considered are ruptures of one or more steam generator tubes leading to water or steam ejection into the primary system or anomalous operating conditions which my cause damage due to excessive thermal stress. Examples are the complete dryout as a consequence of feedwater interrupt in connection with continuing gas heating and the reflooding of the secondary channel with cold feedwater after dryout. The steam generator program which is capable to simulate accidents of this type is written as a module which can be implemented into a program system fur the simulation of the total heat rejection system. It based on an advanced mathematical model for the two phase flow taking deviations from thermal equilibrium into account. Mass, energy and momentum balances for the primary and secondary fluid and the heat diffusion equations for the heat exchanging wall form a system of coupled differential equations which is solved numerically by an algorithm which is stiffly stable and suppresses effectively oscillations of numerical origin. Results of the simulation of transients of the type mentioned above will be presented and discussed. (author)

Open channel steam generator feedwater system

International Nuclear Information System (INIS)

Kim, R.F.; Min-Hsiung Hu.

1985-01-01

A steam generator which utilizes a primary fluid to vaporize a secondary fluid is provided with an open flow channel and elevated discharge nozzle for the introduction of secondary fluid. The discharge nozzle is positioned above a portion of the inlet line such that the secondary fluid passes through a vertical section of inlet line prior to its discharge into the open channel. (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)

Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

Energy Technology Data Exchange (ETDEWEB)

Papp, L. [Inst. of Material Engineering, Ostrava (Switzerland)

1995-12-31

After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed.

Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

Energy Technology Data Exchange (ETDEWEB)

Papp, L [Inst. of Material Engineering, Ostrava (Switzerland)

1996-12-31

After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed.

Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

International Nuclear Information System (INIS)

Papp, L.

1995-01-01

After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed

PACTEL: Experiments on the behaviour of the new horizontal steam generator

Energy Technology Data Exchange (ETDEWEB)

Kouhia, J.; Riikonen, V.; Purhonen, H. [VTT Energy, Lappeenranta (Finland)

1995-12-31

Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

PACTEL: Experiments on the behaviour of the new horizontal steam generator

Energy Technology Data Exchange (ETDEWEB)

Kouhia, J; Riikonen, V; Purhonen, H [VTT Energy, Lappeenranta (Finland)

1996-12-31

Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

Leak suppression at steam generator man-, hand-, and eyeholes

International Nuclear Information System (INIS)

Sylvain, C.; Sutz, P.; Gemma, A.

1988-01-01

Plant unavailability associated with primary and secondary holes is approximately the same as that caused by steam generator tube defects, i.e., 0.5%. Problems encountered with steam generator man-, hand-, and eyeholes during plant operation have led Electricite de France (EdF) and Framatome to improve hole seal design and to develop robots for closing and cleaning them. The data base available in France in this field on some 150 steam generators in 900- and 1300-MW(electric) pressurized water reactors (the equivalent of 300 reactor-yr of operation) has been the base of the developments described in this paper. Incidents occurring in operation primarily concern had-and inspection holes located on the steam generator's secondary side. They include four kinds: (1) leakage detected in operation, requiring forced outages, (2) leakage detected during plant restart after a scheduled shutdown and resulting in a restart delay, (3) pitting of seal mating surfaces, not inducing any leakage but jeopardizing subsequent compliance and requiring difficult and costly repairs, and (4) seizing of screws or bolts. New primary and secondary hole stud tightening and maintenance machines help to improve the efficiency of the in-service closing operations. They provide savings of up to 80% on labor, duration of operations, and exposure

Design and related R and D works of 'Monju' steam generators

Energy Technology Data Exchange (ETDEWEB)

Nakai, Y; Imanaka, N; Hoshi, Y; Tanaka, K; Hori, M; Yoshikawa, Y

1975-07-01

The steam generator is considered to be one of the key components in LMFBR plant. Helical coil type steam generator is selected as a reference for the first Japanese demonstration plant 'MONJU'. The paper gives the structural and functional features of 'MONJU' steam generator together with a brief description of secondary cooling system. The related R and D works are also illustrated. (author)

Development of laser repair technique for cutting impurities in secondary side of a steam generator

Energy Technology Data Exchange (ETDEWEB)

Kim, Cheol Jung; Chung, Chin Man; Kim, Min Suk; Jeong, Tae Moon

1999-12-01

In this research, the laser repair technique was investigated for the purpose of cutting impurities in secondary side of a steam generator. For this research, a high quality Nd:YAG laser was manufactured and the beam delivery experiments was performed with multimode optical fibers. Also, the small size focusing system was designed and fabricated for remote cutting and the laser cutting experiment was performed. (author)

Fission product retention during faults involving steam generator tube rupture

International Nuclear Information System (INIS)

Rodliffe, R.S.

1983-08-01

In some PWR fault conditions, such as stuck open safety relief valve in the secondary circuit or main steam line break, the release of fission products to the atmosphere may be increased by the leakage of primary coolant into the secondary circuit following steam generator tube rupture. The release may be reduced by retention either within the primary circuit or within the affected steam generator unit (SGU). The mechanisms leading to retention are reviewed and quantified where possible. The parameters on which any analysis will be most critically dependent are identified. Fission product iodine and caesium may be retained in the secondary side of a SGU either by partition to retained water or by droplet deposition on surfaces and subsequent evaporation to dryness. Two extreme simplifications are considered: SGU 'dry', i.e. the secondary side is steam filled, and SGU 'wet', i.e. the tube bundle is covered with water. Consideration is given to: the distribution of fission products between gaseous and aerosol forms; mechanisms for droplet formation, deposition and resuspension; fission product retention during droplet or film evaporation primary coolant mixing and droplet scrubbing in a wet SGU; and the performance of moisture separators and steam driers. (author)

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)

The decommissioning of the BR3 steam generator

International Nuclear Information System (INIS)

Denissen, L.

2006-01-01

A steam generator is a crucial component in a PWR (Pressurized Water Reactor). It is the crossing between the primary, contaminated, circuit and the secondary water-steam circuit. The heat from the primary reactor coolant loop is transferred to the secondary side in thousands of small tubes. Due to several problems in the material of those tubes, like SCC (Stress Corrosion Cracking), insufficient control in water chemistry, which can be the cause of tube leakage, more and more steam generators are replaced today. Only in Belgium, already 17 of them are replaced. The old 300 ton heavy SGs are stored at the 2 nuclear power plants of Doel and Tihange . While it was foreseen in the BR3 strategy to dismantle the steam generator (only 30 ton), we took the opportunity to search for a complete package in the decommissioning of a steam generator. The complete management consists of a decontamination of the primary side followed by the complete dismantling. The first step, the decontamination with MEDOC (water box + tube bundle) has already been achieved in 2002. It has led to an important DF (Decontamination Factor) between 100 and 1000 and an important dose rate reduction. This hard chemical decontamination process has been described earlier in the scientific report 2002 (The BR3 steam generator decontamination with the MEDOC process - M. Ponnet). The second step, the complete dismantling of the SG has been executed in 2005. With the BR3 SG, the main goal was to dismantle it in a safe way and to free release a maximum of material. We've used two cutting tools to perform the job: A HPWJC (High Pressure Water Jet Cutting) tool in combination with a hydraulic robot and a water cooled diamond cable. The last technique was done in close collaboration with the external company Husqvarna. It was important to have an idea of the performance, the efficiency of the cable and the quantity and the type of secondary waste

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)

Chemical-cleaning process evaluation: Westinghouse steam generators. Final report

International Nuclear Information System (INIS)

Cleary, W.F.; Gockley, G.B.

1983-04-01

The Steam Generator Owners Group (SGOG)/Electric Power Research Institute (EPRI) Steam Generator Secondary Side Chemical Cleaning Program, under develpment since 1978, has resulted in a generic process for the removal of accumulated corrosion products and tube deposits in the tube support plate crevices. The SGOG/EPRI Project S150-3 was established to obtain an evaluation of the generic process in regard to its applicability to Westinghouse steam generators. The results of the evaluation form the basis for recommendations for transferring the generic process to a plant specific application and identify chemical cleaning corrosion guidelines for the materials in Westinghouse Steam Generators. The results of the evaluation, recommendations for plant-specific applications and corrosion guidelines for chemical cleaning are presented in this report

Embalse steam generators - status in 2009

International Nuclear Information System (INIS)

Luna, P.; Yetisir, M.; Roy, S.; MacEacheron, R.

2009-01-01

The Embalse Nuclear Generating Station (ENGS) is a CANDU 6, a pressurized heavy water plant, with a net capacity of 648 MW. The primary heat transport system at Embalse includes four Steam Generators (SGs) manufactured by Babcock and Wilcox Canada (B and W). These steam generators are vertical recirculating heat exchangers with Incoloy 800 inverted U-tubes and an integral preheater. Embalse SGs performed very well until the late 1990s, when an increase in tube fretting was noticed in the U-bend region. In-service inspection in 2002 and 2004 confirmed that the cause of the tube fretting was flow accelerated corrosion (FAC) damage of scallop bar supports in the U-bend region. The straight leg tube support plates (TSPs) have also been degrading. Degradation was worst at the top support plates, and it was in the form of material loss on the cold leg. The hot leg TSPs were heavily fouled with deposits and flow areas were blocked. Visual inspections and subsequent studies showed that the cause of the TSP degradation was also FAC. The Embalse SGs have carbon steel supports that make them susceptible to FAC. To mitigate the effects of degraded tube support structures, three additional sets of anti-vibration bars were installed in the U-bend regions of all four steam generators in 2004. In 2007, an improved secondary-side chemistry specification was implemented to reduce the FAC rate and the hot leg TSPs was waterlanced. A root cause analysis and condition assessment was performed for the tube supports in 2007. Fitness for Service (FFS) evaluation was completed using the Canadian Industry Guidelines for steam generator tubes. The steam generators were returned to service and the plan has operated without another forced outage to date. The FAC degradation of the carbon steel U-bend tube support systems has had the most significant impact on the plant operation causing a number of forced outages. The discovery of the extent of TSP degradation and difficulties to repair TSPs

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

Steam generator tube rupture effects on a LOCA

International Nuclear Information System (INIS)

LaChance, J.L.

1979-01-01

A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

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.

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

Probabilistic methodology for assessing steam generator tube inspection - Phase II: CANTIA - a probabilistic method for assessing steam generator tube inspections

International Nuclear Information System (INIS)

Harris, J.E.; Gorman, J.A.; Turner, A.P.L.

1997-03-01

The objectives of this project were to develop a computer-based method for probabilistic assessment of inspection strategies for steam generator tubes, and to document the source code and to provide a user's manual for it. The program CANTIA was created to fulfill this objective, and the documentation and verification of the code is provided in this volume. The user's manual for CANTIA is provided as a separate report. CANTIA uses Monte Carlo techniques to determine approximate probabilities of steam generator tube failures under accident conditions and primary-to-secondary leak rates under normal and accident conditions at future points in time. The program also determines approximate future flaw distributions and non-destructive examination results from the input data. The probabilities of failure and leak rates and the future flaw distributions can be influenced by performing inspections of the steam generator tubes at some future points in time, and removing defective tubes from the population. The effect of different inspection and maintenance strategies can therefore be determined as a direct effect on the probability of tube failure and primary-to-secondary leak rate

Chemical cleaning of nuclear (PWR) steam generators

International Nuclear Information System (INIS)

Welty, C.S. Jr.; Mundis, J.A.

1982-01-01

This paper reports on a significant research program sponsored by a group of utilities (the Steam Generator Owners Group), which was undertaken to develop a process to chemically remove corrosion product deposits from the secondary side of pressurized water reactor (PWR) power plant steam generators. Results of this work have defined a process (solvent system and application methods) that is capable of removing sludge and tube-to-tube support plate crevice corrosion products generated during operation with all-volatile treatment (AVT) water chemistry. Considers a plant-specific test program that includes all materials in the steam generator to be cleaned and accounts for the physical locations (proximity and contact) of those materials. Points out that prior to applying the process in an operational unit, the utility, with the participation of the NSSR vendor, must define allowable total corrosion to the materials of construction of the unit

Steam Generator Group Project. Annual report, 1982

International Nuclear Information System (INIS)

Clark, R.A.; Lewis, M.

1984-02-01

The Steam Generator Group Project (SGGP) is an NRC program joined by additional sponsors. The SGGP utilizes a steam generator removed from service at a nuclear plant (Surry 2) as a vehicle for research on a variety of safety and reliability issues. This report is an annual summary of progress of the program for 1982. Information is presented on the Steam Generator Examination Facility (SGEF), especially designed and constructed for this research. Loading of the generator into the SGEF is then discussed. The report then presents radiological field mapping results and personnel exposure monitoring. This is followed by information on field reduction achieved by channel head decontaminations. The report then presents results of a secondary side examination through shell penetrations placed prior to transport, confirming no change in generator condition due to transport. Decontamination of the channel head is discussed followed by plans for eddy current testing and removal of the plugs placed during service. Results of a preliminary profilometry examination are then provided

Functional performance of the helical coil steam generator, Consolidated Nuclear Steam Generator (CNSG) IV system. Executive summary report

International Nuclear Information System (INIS)

Watson, G.B.

1975-10-01

The objective of this project was to study the functional performance of the CNSG - IV helical steam generator to demonstrate that the generator meets steady-state and transient thermal-hydraulic performance specifications and that secondary flow instability will not be a problem. Economic success of the CNSG concepts depends to a great extent on minimizing the size of the steam generator and the reactor vessel for ship installation. Also, for marine application the system must meet stringent specifications for operating stability, transient response, and control. The full-size two-tube experimental unit differed from the CNSG only in the number of tubes and the mode of primary flow. In general, the functional performance test demonstrated that the helical steam generator concept will exceed the specified superheat of 35F at 100% load. The experimental measured superheat at comparable operating conditions was 95F. Testing also revealed that available computer codes accurately predict trends and overall performance characteristics

Decontamination of Steam Generator tube using Abrasive Blasting Technology

International Nuclear Information System (INIS)

Min, B. Y.; Kim, G. N.; Choi, W. K.; Lee, K. W.; Kim, D. H.; Kim, K. H.; Kim, B. T.

2010-01-01

As a part of a technology development of volume reduction and self disposal for large metal waste project, We at KAERI and our Sunkwang Atomic Energy Safety (KAES) subcontractor colleagues are demonstrating radioactively contaminated steam generator tube by abrasive blasting technology at Kori-1 NPP. A steam generator is a crucial component in a PWR (pressurized Water Reactor). It is the crossing between the primary, contaminated, circuit and the secondary waste-steam circuit. The heat from the primary reactor coolant loop is transferred to the secondary side in thousands of small tubes. Due to several problems in the material of those tube, like SCC (Stress Corrosion Cracking), insufficient control in water chemistry, which can be cause of tube leakage, more and more steam generators are replaced today. Only in Korea, already 2 of them are replaced and will be replaced in the near future. The retired 300 ton heavy Steam generator was stored at the storage waste building of Kori NPP site. The steam generator waste has a large volume, so that it is necessary to reduce its volume by decontamination. A waste reduction effect can be obtained through decontamination of the inner surface of a steam generator. Therefore, it is necessary to develop an optimum method for decontamination of the inner surface of bundle tubes. The dry abrasive blasting is a very interesting technology for the realization of three-dimensional microstructures in brittle materials like glass or silicon. Dry abrasive blasting is applicable to most surface materials except those that might be shattered by the abrasive. It is most effective on flat surface and because the abrasive is sprayed and can also applicable on 'hard to reach' areas such as inner tube ceilings or behind equipment. Abrasive decontamination techniques have been applied in several countries, including Belgium, the CIS, France, Germany, Japan, the UK and the USA

Steam generator assessment for sustainable power plant operation

International Nuclear Information System (INIS)

Drexler, Andreas; Fandrich, Joerg; Ramminger, Ute; Montaner-Garcia, Violeta

2012-09-01

extensive range of input-data and it is therefore not comparable with a basic questionnaire. It provides the means for monitoring steam generator condition comparing various plant data and, in combination with lifetime management tools, verifying statements on the steam generator performance. AREVA Fouling Index Tool Box was designed, evaluated and validated with Siemens designed steam generators and High-AVT secondary side chemistry treatment. Based on the gathered excellent experience AREVA applied the Fouling Index tool box to other steam generator models with different balance-of-plant designs, materials and secondary side water chemistry treatments. Application examples are presented in this paper. (authors)

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)

Condition monitoring of steam generator by estimating the overall heat transfer coefficient

International Nuclear Information System (INIS)

Furusawa, Hiroaki; Gofuku, Akio

2013-01-01

This study develops a technique for monitoring in on-line the state of the steam generator of the fast-breeder reactor (FBR) “Monju”. Because the FBR uses liquid sodium as coolant, it is necessary to handle liquid sodium with caution due to its chemical characteristics. The steam generator generates steam by the heat of secondary sodium coolant. The sodium-water reaction may happen if a pinhole or crack occurs at the thin metal tube wall that separates the secondary sodium coolant and water/steam. Therefore, it is very important to detect an anomaly of the wall of heat transfer tubes at an early stage. This study aims at developing an on-line condition monitoring technique of the steam generator by estimating overall heat transfer coefficient from process signals. This paper describes simplified mathematical models of superheater and evaporator to estimate the overall heat transfer coefficient and a technique to diagnose the state of the steam generator. The applicability of the technique is confirmed by several estimations using simulated process signals with artificial noises. The results of the estimations show that the developed technique can detect the occurrence of an anomaly. (author)

Large-leak sodium-water reaction analysis for steam generators

International Nuclear Information System (INIS)

Sakano, K.; Shindo, Y.; Hori, M.

1975-01-01

The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

Large-leak sodium-water reaction analysis for steam generators

Energy Technology Data Exchange (ETDEWEB)

Sakano, K; Shindo, Y; Hori, M

1975-07-01

The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

Assesment of integrity of WWER steam generator tubes

International Nuclear Information System (INIS)

Splichal, K.; Brozova, A.; Zdarek, J.

1992-01-01

Full text: The leak rates measurement project was held to give experimental data enabling the Czechoslovak Atomic Agency Inspection to decree the change in the Technical Specification allowable limit of steam generator activity release on secondary side. The WWER types of nuclear power plants in Czechoslovakia have horizontal steam generators. The tubes studying in frame of the project belong to steam generator WWER- 440 type, the diameter of tube is 16 mm, the wall thickness 1.4 mm. The subject of the project was the measurement of service leak rates of typical in service cracks. Secondary side stress corrosion cracks were determined as the typical crack created in service condition. These cracks were prepared in tubes artificially by exposition in chloride environment accompanied by an internal stress. The experimental device consisted of a pressure vessel connected with pressure water loop, a cooling vessel for leakage medium and a measuring vessel. The leak rates were determined as a slope of plots the leakage volume - time. Inside the pressure vessel the steam generator operation environment was simulated. It means: primary side of tube 12.5 MPa, Z90 deg. C, secondary side -4.6MPa, 250 deg. C, water service quality. We observed reduce of leak rate in course of time in each experiment. We suppose the tubes were stopped up by deposits formed in manufacturing of crack and in experiment. Our opinion has been proved by fractography. Project results in recommendation for in service leak rate limit based on safety factors with respect to critical crack lengths and for determination of tube plugging criteria. (author)

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

Procedure for generating steam and steam generator for operating said procedure

International Nuclear Information System (INIS)

Chlique, Bernard.

1975-01-01

This invention concerns the generation of steam by bringing the water to be vaporised into indirect thermal exchange relation with the heating steam which condenses when passing in series, along alternate routes, through bundles of tubes immersed in a vaporising chamber. A number of steam generators working on this principle already exist. The purpose of the invention is to modify the operating method of these steam generators by means of a special disposition making it possible to build a compact unit including an additional bundle of tubes heated by the condensates collected at the outlet of each bundle through which the heating steam passes [fr

Steam generator tubesheet waterlancing at Bruce B

Energy Technology Data Exchange (ETDEWEB)

Persad, R. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada); Eybergen, D. [Bruce Power, Tiverton, Ontario (Canada)

2006-07-01

High pressure water cleaning of steam generator secondary side tubesheet surfaces is an important and effective strategy for reducing or eliminating under-deposit chemical attack of the tubing. At the Bruce B station, reaching the interior of the tube bundle with a high-pressure water lance is particularly challenging due to the requirement to setup on-boiler equipment within the containment bellows. This paper presents how these and other design constraints were solved with new equipment. Also discussed is the application of new high-resolution inter-tube video probe capability to the Bruce B steam generator tubesheets. (author)

A study on integrity of LMFBR secondary cooling system to hypothetical tube failure propagation in the steam generator

International Nuclear Information System (INIS)

Yoshihisa Shindo; Kazuo Haga

2005-01-01

Full text of publication follows: A fundamental safety issue of liquid-metal-cooled fast breeder reactor (LMFBR) is to maintain the integrity of the secondary cooling system components against violent chemical sodium-water reaction caused by the water leak from the heat transfer tube of steam generators (SG). The produced sodium-water reaction jet would attack more severely surrounding tubes and would cause other tube failures (tube failure propagation), if it was assumed that the water leak was not detected by function-less detectors and proper operating actions to mitigate the tube failure propagation, such as isolations of the SG from the secondary cooling system and turbine water/steam system, and blowing water and steam inside tubes in the SG, were not taken. This study has been made focusing on the affection of large-scale water leak enlarged due to SG tube failure propagation to the structural integrity of the secondary cooling system because the generated pressure pulse caused by a large-scale sodium-water reaction might break heat transfer tubes of the intermediate heat exchanger (IHX). The present work has been made as one part of the study of probabilistic safety assessment (PSA) of LMFBR, because if the heat-transfer tubes of IHX were failed, the reactor core may be affected by the pressure pulse and/or by the sodium-water reaction products transported through the primary cooling system. As tools for PSA of the water leak incident of SG, we have developed QUARK-LP Version 4 code that mainly analyzes the high temperature rupture phenomena and estimates the number of failed tubes during the middle-scale water leak. The pressure pulse behavior generated by sodium-water reaction in the failure SG and the pressure propagation in the secondary cooling system are calculated by using the SWAAM-2 code developed by ANL. Furthermore, the quasi-steady state high pressure and temperature of the secondary cooling system in a long term is estimated by using the SWAAM

Dismantling of the 50 MW steam generator test facility

International Nuclear Information System (INIS)

Nakai, S.; Onojima, T.; Yamamoto, S.; Akai, M.; Isozaki, T.; Gunji, M.; Yatabe, T.

1997-01-01

We have been dismantling the 50MW Steam Generator Test Facility (50MWSGTF). The objectives of the dismantling are reuse of sodium components to a planned large scale thermal hydraulics sodium test facility and the material examination of component that have been operated for long time in sodium. The facility consisted of primary sodium loop with sodium heater by gas burner as heat source instead of reactor, secondary sodium loop with auxiliary cooling system (ACS) and water/steam system with steam temperature and pressure reducer instead of turbine. It simulated the 1 loop of the Monju cooling system. The rated power of the facility was 50MWt and it was about 1/5 of the Monju power plant. Several sodium removal methods are applied. As for the components to be dismantled such as piping, intermediate heat exchanger (IHX), air cooled heat exchangers (AC), sodium is removed by steam with nitrogen gas in the air or sodium is burned in the air. As for steam generators which material tests are planned, sodium is removed by steam injection with nitrogen gas to the steam generator. The steam generator vessel is filled with nitrogen and no air in the steam generator during sodium removal. As for sodium pumps, pump internal structure is pulled out from the casing and installed into the tank. After the installation, sodium is removed by the same method of steam generator. As for relatively small reuse components such as sodium valves, electromagnet flow meters (EMFs) etc., sodium is removed by alcohol process. (author)

Steam generator tube performance

International Nuclear Information System (INIS)

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

1984-10-01

A review of the performance of steam generator tubes in 116 water-cooled nuclear power reactors showed that tubes were plugged at 54 (46 percent) of the reactors. The number of tubes removed from service decreased from 4 692 (0.30 percent) in 1981 to 3 222 (0.20 percent) in 1982. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that have used only volatile treatment, with or without condensate demineralization

Transient analysis of a U-tube natural circulation steam generator

Energy Technology Data Exchange (ETDEWEB)

Gaikwad, A J; Kumar, Rajesh; Bhadra, Anu; Chakraborty, G; Venkat Raj, V [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

A computer code has been developed, for transient thermal-hydraulic analysis of proposed 500 MWe PHWR steam generator. The transient behaviour of a nuclear power plant is very much dependent on the steam generator performance, as it provides a thermal linkage between the primary and secondary systems. Study of dynamics of steam generator is essential for over all power plant dynamics as well as design of control systems for steam generator. A mathematical model has been developed for the simulation of thermal-hydraulic phenomena in a U tube natural circulation steam generator. Fluid model is based on one dimensional, nonlinear, single fluid conservation equations of mass, momentum, energy and equation of state. This model includes coupled two phase flow heat transfer and natural circulation. The model accounts for both compressibility and thermal expansion effects. The process simulation and results obtained for transients such as step change in load and total loss of feed water are presented. (author). 5 refs., 7 figs.

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)

New Technology for Corrosion Mitigation of Steam Generator Tubesheet in Secondary Side Environments

International Nuclear Information System (INIS)

Hur, Do Haeng; Choi, Myung Sik; Lee, Deok Hyun; Han, Jung Ho

2013-01-01

Denting has been mitigated by a modification of the design and material of the tube support structures, it has been an inevitable problem in the crevice region of the top of the tubesheet(TTS). Denting at the TTS has been a significant concern regardless of the tube materials. This is because it is a mechanical process resulted from a volume expansion of corrosion products of the tubesheet materials. It should be noted that the corrosion rate of low alloy tubesheet materials is accelerated due to the presence of corrosion products accumulated at the top of the tubesheet. Therefore a reduction of the corrosion rate of the tubesheet material should be a key strategy to prevent tube denting at the TTS as well as an improvement of the secondary water chemistry. This paper provides a new technology to prevent denting by cladding the secondary side surface of the tubesheet with a corrosion resistant material. In this study, Alloy 690 material on the surface of the SA508 tubesheet was cladded to a thickness of about 9mm. The corrosion rates of the SA508 original tubesheet and Alloy 690 clad material were measured in acidic and caustic simulated environments. Denting has been a precursor of stress corrosion cracking in nuclear steam generator tubing, although it may be mitigated by a design and material modification of the tube support structures and secondary water chemistry control. Corrosion resistant Alloy 690 tubing is not an exception because denting at the TTS is due to corrosion of the tubesheet material. In this paper, a new technology was suggested to prevent denting at the TTS by cladding the secondary side surface of the tubesheet with a corrosion resistant material. It was verified that the corrosion rates of a tubesheet with an Alloy 690 clad layer drastically decreased in both acidic and alkaline environments, even inside the magnetite sludge pile. Because the cladding processes of Alloy 690 have already been applied to the primary side surface of a

Estimating probable flaw distributions in PWR steam generator tubes

International Nuclear Information System (INIS)

Gorman, J.A.; Turner, A.P.L.

1997-01-01

This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses

Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

Energy Technology Data Exchange (ETDEWEB)

Fynan, Douglas A.; Ahn, Kwang-Il, E-mail: kiahn@kaeri.re.kr

2016-12-15

Highlights: • Pressure drop-flow rate curves for superheated steam in U-tubes were generated. • Forward flow of hot steam is favored in the longer and taller U-tubes. • Reverse flow of cold steam is favored in short U-tubes. • Steam generator U-tube bundle geometry and tube diameter are important. • Need for correlation development for natural convention heat transfer coefficient. - Abstract: Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.

Corrosion Processes of the CANDU Steam Generator Materials in the Presence of Silicon Compounds

International Nuclear Information System (INIS)

Lucan, Dumitra; Fulger, Manuela; Velciu, Lucian; Lucan, Georgiana; Jinescu, Gheorghita

2006-01-01

The feedwater that enters the steam generators (SG) under normal operating conditions is extremely pure but, however, it contains low levels (generally in the μg/l concentration range) of impurities such as iron, chloride, sulphate, silicate, etc. When water is converted into steam and exits the steam generator, the non-volatile impurities are left behind. As a result of their concentration, the bulk steam generator water is considerably higher than the one in the feedwater. Nevertheless, the concentrations of corrosive impurities are in general sufficiently low so that the bulk water is not significantly aggressive towards steam generator materials. The impurities and corrosion products existing in the steam generator concentrate in the porous deposits on the steam generator tubesheet. The chemical reactions that take place between the components of concentrated solutions generate an aggressive environment. The presence of this environment and of the tubesheet crevices lead to localized corrosion and thus the same tubes cannot ensure the heat transfer between the fluids of the primary and secondary circuits. Thus, it becomes necessary the understanding of the corrosion process that develops into SG secondary side. The purpose of this paper is the assessment of corrosion behavior of the tubes materials (Incoloy-800) at the normal secondary circuit parameters (temperature = 2600 deg C, pressure = 5.1 MPa). The testing environment was demineralized water containing silicon compounds, at a pH=9.5 regulated with morpholine and cyclohexyl-amine (all volatile treatment - AVT). The paper presents the results of metallographic examinations as well as the results of electrochemical measurements. (authors)

Thermal-Hydraulic Design of the Modular Once Through Helical Steam Generator

International Nuclear Information System (INIS)

Mazufri, C.M

2000-01-01

The steam generator system of the CAREM reactor consists of twelve individual modules located in the annular place between the pressure vessel and barrel walls. Each steam generator module consists of a tube system, an upper header, an external shroud, a collector and a lower seal.The tube system is an arrangement of several multi-start cylindrical coils.In the present work the computation of the necessary heat transfer area to fulfill the heat removal requirements from the primary circuit, and the pressure drop in the primary and secondary side of the helical design of a modular steam generator is presented. Additionally, a first order estimation of the restriction to be used in the secondary side to assure the thermal-hydraulic stability is also made.It is concluded that an array of 6 concentric cylindrical coils fulfills the necessary design requirements

Steam generator inspection activities at the EPRI NDE Center

International Nuclear Information System (INIS)

Krzywosz, K.

1988-01-01

Various types of corrosion and mechanical damage continue to affect the availability of both recirculating and once-through steam generators. Both the tube bundle and its supporting structure are affected. Intergranular attack and stress corrosion cracking (SCC) are the corrosion-assisted tube-wall damage mechanisms of most concern at this time. Fatigue cracking and fretting at antivibration bars are currently the mechanical damage forms causing most concern. Improved NDE equipment and techniques are providing better detection and characterization of adverse conditions within the steam generators and doing it at an earlier stage. This allows timely corrective action. To maintain the projected life expectancy of existing and new steam generators, remedial measures have been implemented. These measures include shot- or roto-peening, U-bend stress relief, chemical cleaning of secondary side, and sleeving of tubes. The improved NDE technology will also be instrumental in monitoring and assessing the effectiveness of the remedial measures. The revision of guidance documents for steam generator in-service inspection (ISI) is providing more relevant information to support this complex operation. A multitasked project is described that includes evaluation of steam generator tube NDE technology, transfer of this technology to utilities, and rapid response utility assistance

Visual inspection equipment for the secondary side of steam generators. Final report

International Nuclear Information System (INIS)

Hinton, M.S.; Chagnon, C.W.

1981-05-01

The development of miniature visual examination equipment capable of inspecting regions inside the tube bundles of nuclear steam generators to determine such factors as gap condition, tube outside surface condition, and material buildup is described. Tube-to-support plate interface regions at the tenth-support plate level in a once-through steam generator (Oconee 1) were visually examined. Other areas in the peripheral region were examined for conditions associated with available eddy-current and profilometer data

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

Design of jet manipulator for sludge lancing for steam generators

International Nuclear Information System (INIS)

Kumar, Kundan; Nathani, D.K.; Kayal, J.N.; Rupani, B.B.

2006-01-01

The sludge accumulation in secondary side of mushroom type steam generators of Indian Pressurised Heavy Water Reactors (PHWRs) may lead to loss of thermal efficiency and corrosion. Sludge removal is required to minimise such effects for safe and enhanced operating life of the steam generators. A sludge lancing system has been developed for sludge removal from the secondary side of the steam generators. Jet Manipulator is one of the various modules of the sludge lancing system. The JM consists of three modules namely walker, elevator and nozzle heads. Each module is designed to pass through hand hole, having 180 mm diameter and 100 mm wide gap between steam generator shell and shroud. These three modules are connected to each other by quick connecting type joints and are having their specific functions. The walker crawls by step of single pitch of the tube along the central no-tube lane of the steam generator by taking lateral supports on the nearest tubes. The elevator is capable of lifting the nozzle head to a suitable height required for lancing operation of entire tube sheet of the steam generator. The nozzle head directs the multiple jets along the narrow inter tube lanes having 3 mm width, on both sides of the central no-tube lane. The nozzle can be set to move at different elevations such that the multiple jets will graze along the narrow tube lane to create the sludge lancing action. The provision exists for movement of JM in both directions, i.e. forward and reverse. This paper highlights the objective, design and development, selection of nozzles, qualification and performance evaluation of JM. The manipulator is remotely operable by compressed air in the forward and reverse direction in the central no-tube lane to position the nozzle head in the horizontal direction. (author)

GVTRAN-PC-a steam generator transient simulator

International Nuclear Information System (INIS)

Nakata, H.

1991-02-01

Since an accuracy and inexpensive analysis capability is one of the desirable requirement in the reactor licensing procedure and, also, in instances when a severe accident sequence and its degree of severity must be estimated, the present work tries partially to fulfill that present need by developing a fast and acceptably accurate simulator. The present report presents the methodology utilized to develop GVTRAN-PC program, the steam generator simulation program for the microcomputer environment, which possess a capability to reproduce the experimental data with accuracies comparable to those of the mainframe simulators. The methodology is based on the mass and energy conservation in the control volumes which represents both the primary and the secondary fluid in the U-tube steam generator. The quasi-static momentum conservation in the secondary fluid control volumes determines in a semi-iterative scheme the liquid level in the feedwater chamber. The implementation of the moving boundary technique has allowed the tracking of the boundary of bulk boiling region with the utilization of a reduced number of control volumes in the tube region. GVTRAN-PC program has been tested against typical PWR pump trip transient experimental data and the calculation results showed good agreement in most representative parameters, viz. the feedchamber water-level and the steam dome pressure. (author)

CFD evaluation on the thermohydraulic characteristics of tube support plates in steam generator

Energy Technology Data Exchange (ETDEWEB)

Zhang, B.; Zhang, H.; Han, B.; Yang, B.W. [Xi' an Jiaotong Univ. (China). School of Nuclear Science and Technology; Mo, S.J.; Ren, H.B.; Qin, J.M.; Zuo, C.P. [China Nuclear Power Design Co. Ltd., ShenZhen (China)

2016-07-15

The integrity and thermal hydraulic characteristics of steam generator are of great concern in the nuclear industry. The tube support plates (TSP), one of the most important components of the steam generator, not only support the heat transfer tubes, but also affect the flow dynamic and thermal hydraulic characteristics of the secondary-side flow inside the steam generator. Different working conditions, ranging from single-phase adiabatic condition to two-phase high-void boiling condition, are simulated and analyzed. Calculated void fraction, under simple geometry, agrees well with the experiment data whilst the simulated heat transfer coefficient is tremendously close to the empirical correlation. Temperature, void fraction, and velocity distributions in different locations show reasonable distribution. The simulation results indicate that TSP can enhance the heat transfer in the secondary side of the steam generator. On the top of TSP, with the increase in cross-section flow area, the back-flow phenomenon occurs, which might lead to the contamination of precipitation.

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

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 generators of Phenix: Measurement of the hydrogen concentration in sodium for detecting water leaks in the steam generator tubes

International Nuclear Information System (INIS)

Cambillard, E.; Lacroix, A.; Langlois, J.; Viala, J.

1975-01-01

The Phenix secondary circuits are provided with measurement systems of hydrogen concentration in sodium, that allow for the detection of possible water leaks in steam generators and the location of a faulty module. A measurement device consists of : a detector with nickel membranes of 0, 3 mm wall thickness, an ion pump with a 200 l/s flow rate, a quadrupole mass spectrometer and a calibrated hydrogen leak. The temperature correction is made automatically. The main tests carried out on the leak detection systems are reported. Since the first system operation (October 24, 1973), the measurements allowed us to obtain the hydrogen diffusion rates through the steam generator tube walls. (author)

Steam generators secondary side chemical cleaning at Point Lepreau using the Siemens high temperature process

International Nuclear Information System (INIS)

Verma, K.; MacNeil, C.; Odar, S.; Kuhnke, K.

1997-01-01

This paper describes the chemical cleaning of the four steam generators at the Point Lepreau facility, which was accomplished as a part of a normal service outage. The steam generators had been in service for twelve years. Sludge samples showed the main elements were Fe, P and Na, with minor amounts of Ca, Mg, Mn, Cr, Zn, Cl, Cu, Ni, Ti, Si, and Pb, 90% in the form of Magnetite, substantial phosphate, and trace amounts of silicates. The steam generators were experiencing partial blockage of broached holes in the TSPs, and corrosion on tube ODs in the form of pitting and wastage. In addition heat transfer was clearly deteriorating. More than 1000 kg of magnetite and 124 kg of salts were removed from the four steam generators

Steam generators - problems and prognosis

International Nuclear Information System (INIS)

Tapping, R.L.

1997-05-01

Steam-generator problems, largely a consequence of corrosion and fouling, have resulted in increased inspection requirements and more regulatory attention to steam-generator integrity. In addition, utilities have had to develop steam-generator life-management strategies, including cleaning and replacement, to achieve design life. This paper summarizes the pertinent data to 1993/1994, and presents an overview of current steam-generator management practices. (author)

Steam generator tube performance

International Nuclear Information System (INIS)

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

1983-08-01

A review of the performance of steam generator tubes in 110 water-cooled nuclear power reactors showed that tubes were plugged at 46 (42 percent) of the reactors. The number of tubes removed from service increased from 1900 (0.14 percent) in 1980 to 4692 (0.30 percent) in 1981. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that used all-volatile treatment since start-up. At one reactor a large number of degraded tubes were repaired by sleeving which is expected to become an important method of tube repair in the future

EPRI steam generator programs

International Nuclear Information System (INIS)

Martel, L.J.; Passell, T.O.; Bryant, P.E.C.; Rentler, R.M.

1977-01-01

The paper describes the current overall EPRI steam generator program plan and some of the ongoing projects. Because of the recent occurrence of a corrosion phenomenon called ''denting,'' which has affected a number of operating utilities, an expanded program plan is being developed which addresses the broad and urgent needs required to achieve improved steam generator reliability. The goal of improved steam generator reliability will require advances in various technologies and also a management philosophy that encourages conscientious efforts to apply the improved technologies to the design, procurement, and operation of plant systems and components that affect the full life reliability of steam generators

Materials performance in operating PWR steam generators

International Nuclear Information System (INIS)

Weeks, J.R.

1975-01-01

The Inconel-600 tubing in operating PWR steam generators has developed leaks due to intergranular stress corrosion cracking or a general wastage attack, originating from the secondary side of the tubing. Corrosion has been limited to those areas of the steam generators where limited coolant circulation and high heat flux have caused impurities to concentrate. Wastage or pitting attack has always been associated with local concentration of sodium hydrogen phosphates, whereas stress corrosion has been associated with local concentration of sodium or potassium hydroxides. The only instance of stress corrosion originating from the primary side occurred on cold-worked tubing when hydrogen was not added to getter oxygen, and LiOH was not added to raise the pH of the primary coolant. All PWR manufacturers are now recommending that the phosphate treatment of the secondary coolant be abandoned in favor of an all-volatile treatment. Experience in operating plants has shown, however, that removal of phosphate-rich sludge deposits is difficult, and that further wastage and/or intergranular stress corrosion may develop; the residual sodium phosphates gradually convert by reaction with corrosion product hydroxides to sodium hydroxide, which remains concentrated in the limited flow areas. Improvements in circulation patterns have been achieved by inserting flow baffles in some PWR steam generators. Inservice monitoring by eddy current techniques is useful for detecting corrosion-induced defects in the tubing, but irreproducibility in field examinations can lead to uncertainties interpreting the results. (U.S.)

Design of SMART steam generator cassette

International Nuclear Information System (INIS)

Kim, Y. W.; Kim, J. I.; Jang, M. H.

2001-01-01

Basic design development for the steam generator to be installed in the integral reactor SMART has been performed. Optimization of the steam generator shape, determination of the basic dimension and confirmation of the structural strength have been carried out. Individual steam generator cassette can be replaced in the optimized design concept of steam generator. Shape design of the steam generator cassette has been done on the computer based on 3-D CAE strategy. The structural integrity of the developed steam generator was investigated by performing the dynamic analysis for the steam generator cassette, flow induced vibration analysis for the tube bundle, and the thermo-mechanical analysis for the module header and tube. As for the manufacturing of steam generator, the numerical and the experimental simulation have been carried to control the amount of spring back and to eliminate residual stress. SMART steam generator cassette was developed by a sequential research of the aforementioned activities

Steam Generator Group Project. Task 6. Channel head decontamination

International Nuclear Information System (INIS)

Allen, R.P.; Clark, R.L.; Reece, W.D.

1984-08-01

The Steam Generator Group Project utilizes a retired-from-service pressurized-water-reactor steam generator as a test bed and source of specimens for research. An important preparatory step to primary side research activities was reduction of the radiation field in the steam generator channel head. This task report describes the channel head decontamination activities. Though not a programmatic research objective it was judged beneficial to explore the use of dilute reagent chemical decontamination techniques. These techniques presented potential for reduced personnel exposure and reduced secondary radwaste generation, over currently used abrasive blasting techniques. Two techniques with extensive laboratory research and vendors prepared to offer commercial application were tested, one on either side of the channel head. As indicated in the report, both techniques accomplished similar decontamination objectives. Neither technique damaged the generator channel head or tubing materials, as applied. This report provides details of the decontamination operations. Application system and operating conditions are described

Modelling horizontal steam generator with ATHLET. Verification of different nodalization schemes and implementation of verified constitutive equations

Energy Technology Data Exchange (ETDEWEB)

Beliaev, J.; Trunov, N.; Tschekin, I. [OKB Gidropress (Russian Federation); Luther, W. [GRS Garching (Germany); Spolitak, S. [RNC-KI (Russian Federation)

1995-12-31

Currently the ATHLET code is widely applied for modelling of several Power Plants of WWER type with horizontal steam generators. A main drawback of all these applications is the insufficient verification of the models for the steam generator. This paper presents the nodalization schemes for the secondary side of the steam generator, the results of stationary calculations, and preliminary comparisons to experimental data. The consideration of circulation in the water inventory of the secondary side is proved to be necessary. (orig.). 3 refs.

Modelling horizontal steam generator with ATHLET. Verification of different nodalization schemes and implementation of verified constitutive equations

Energy Technology Data Exchange (ETDEWEB)

Beliaev, J; Trunov, N; Tschekin, I [OKB Gidropress (Russian Federation); Luther, W [GRS Garching (Germany); Spolitak, S [RNC-KI (Russian Federation)

1996-12-31

Currently the ATHLET code is widely applied for modelling of several Power Plants of WWER type with horizontal steam generators. A main drawback of all these applications is the insufficient verification of the models for the steam generator. This paper presents the nodalization schemes for the secondary side of the steam generator, the results of stationary calculations, and preliminary comparisons to experimental data. The consideration of circulation in the water inventory of the secondary side is proved to be necessary. (orig.). 3 refs.

Processing method for chemical cleaning liquid on the secondary side of steam generator

International Nuclear Information System (INIS)

Nishihara, Yukio; Inagaki, Yuzo.

1993-01-01

Upon processing nitrilotriacetate (NTA), Fe liquid wastes mainly comprising Fe and Cu liquid wastes mainly comprising ethylene diamine and Cu generated upon chemical cleaning on the secondary side of a steam generator, pH of the Fe liquid wastes is lowered to deposit and separate NTA. Then, Fe ions in a filtrates are deposited on a cathode by electrolysis, as well as remaining NTA is decomposed by oxidation at an anode by O 2 gas. Cu liquid wastes are reacted with naphthalene disulfate and Ba ions and the reaction products are separated by deposition as sludges. Remaining Cu ions in the filtrates are deposited on the cathode by electrolysis. With such procedures, concentration of COD(NTA), Fe ions and Cu ions can greatly be reduced. Further, since capacity of the device can easily be increased in this method, a great amount of liquid wastes can be processed in a relatively short period of time. (T.M.)

Modelling the horizontal steam generator with APROS

Energy Technology Data Exchange (ETDEWEB)

Ylijoki, J. [VTT Energy, Espoo (Finland); Palsinajaervi, C.; Porkholm, K. [IVO International Ltd, Vantaa (Finland)

1995-12-31

In this paper the capability of the five- and six-equation models of the simulation code APROS to simulate the behaviour of the horizontal steam generator is discussed. Different nodalizations are used in the modelling and the results of the stationary state runs are compared. Exactly the same nodalizations have been created for the five- and six-equation models. The main simulation results studied in this paper are void fraction and mass flow distributions in the secondary side of the steam generator. It was found that quite a large number of simulation volumes is required to simulate the distributions with a reasonable accuracy. The simulation results of the different models are presented and their validity is discussed. (orig.). 4 refs.

Modelling the horizontal steam generator with APROS

Energy Technology Data Exchange (ETDEWEB)

Ylijoki, J [VTT Energy, Espoo (Finland); Palsinajaervi, C; Porkholm, K [IVO International Ltd, Vantaa (Finland)

1996-12-31

In this paper the capability of the five- and six-equation models of the simulation code APROS to simulate the behaviour of the horizontal steam generator is discussed. Different nodalizations are used in the modelling and the results of the stationary state runs are compared. Exactly the same nodalizations have been created for the five- and six-equation models. The main simulation results studied in this paper are void fraction and mass flow distributions in the secondary side of the steam generator. It was found that quite a large number of simulation volumes is required to simulate the distributions with a reasonable accuracy. The simulation results of the different models are presented and their validity is discussed. (orig.). 4 refs.

North Anna Power Station - Unit 1: Overview of steam generator replacement project activities

International Nuclear Information System (INIS)

Gettler, M.W.; Bayer, R.K.; Lippard, D.W.

1993-01-01

The original steam generators at Virginia Electric and Power Company's (Virginia Power) North Anna Power Station (NAPS) Unit 1 have experienced corrosion-related degradation that require periodic inspection and plugging of steam generator tubes to ensure their continued safe and reliable operation. Despite improvements in secondary water chemistry, continued tube degradation in the steam generators necessitated the removal from service of approximately 20.3 percent of the tubes by plugging, (18.6, 17.3, and 25.1 for steam generators A, B, and C, respectively). Additionally, the unit power was limited to 95 % during, its last cycle of operation. Projections of industry and Virginia Power experience indicated the possibility of mid-cycle inspections and reductions in unit power. Therefore, economic considerations led to the decision to repair the steam generators (i.e., replace the steam generator lower assemblies). Three new Model 51F Steam Generator lower assembly units were ordered from Westinghouse. Virginia Power contracted Bechtel Power Corporation to provide the engineering and construction support to repair the Unit 1 steam generators. On January 4, 1993, after an extended coastdown period, North Anna Unit 1 was brought off-line and the 110 day (breaker-to-breaker) Steam Generator Replacement Project (SGRP) outage began. As of this paper, the outage is still in progress

Sludge cleaning in the steam generators: sludge Lancing e IBL

International Nuclear Information System (INIS)

Montoro, E.; Gonzalez, S.; Calderon, N.

2013-01-01

IBERDROLA Engineering and Construction has echoed the need for plants to remove oxide deposits (sludge) located on the secondary side, on the bottom plate and into the tube bundle steam steam generators. Therefore, and with its partner SAVAC SRA has developed a specific system consisting of applying a capillary water at very high pressure applied directly to the location of these oxides. (Author)

Steam generators, turbines, and condensers. Volume six

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume six covers steam generators (How steam is generated, steam generation in a PWR, vertical U-tube steam generators, once-through steam generators, how much steam do steam generators make?), turbines (basic turbine principles, impulse turbines, reaction turbines, turbine stages, turbine arrangements, turbine steam flow, steam admission to turbines, turbine seals and supports, turbine oil system, generators), and condensers (need for condensers, basic condenser principles, condenser arrangements, heat transfer in condensers, air removal from condensers, circulating water system, heat loss to the circulating water system, factors affecting condenser performance, condenser auxiliaries)

Corrosion Product Measurements to ensure integrity of the Steam Generators in Beznau NPP

International Nuclear Information System (INIS)

Mailand, Irene; Franz, Patrick; Venz, Hartmut

2012-09-01

The Nuclear Power Plant Beznau comprises two identical 380 MWe PWR units with two loops each, commissioned in 1969 and 1971. Westinghouse was responsible for the primary part of the plant and BBC/ABB for the secondary circuit. The original materials used in the secondary systems were made of several copper-based alloys, such as for the Condensers, the Low Pressure Pre-heaters and the Moisture Separator Re-heater. The original Steam Generator Tubes were made of Inconel 600 MA. Regarding its age, the NPP Beznau has to be qualified as an old plant. However, in fact particularly in the last 20 years the plant has undergone an extensive modernisation programme in which about 1.5 billion Swiss Francs have been invested. Important measures were the replacements of the Steam Generators with tubes comprising Inconel 690 TT which was realized at unit 1 in 1993 and at unit 2 in 1999. Copper was completely banished from the secondary system and replaced by stainless and chromium steel. The Condensers were fitted with titanium tubes. The secondary water chemistry had to be changed by these replacements and moved step by step from Low-AVT with a pH of about 9.3 to High-AVT with a pH of 9.8 to 9.9, currently. To ensure the integrity of the new Steam Generators as well as of the whole Secondary System a corrosion product programme was introduced at the end of the Nineties. Several investigations which are performed periodically are represented by analyses of corrosion products, measurements of sludge mass and composition in the Steam Generators, Hide-Out-Return- and mass balance measurements of corrosion products in the whole circuit. Objectives of these investigations are assessments of the efficiency of the water chemistry and trend considerations regarding to the transport of corrosion products and pollutants into the Steam Generator, as well as of the potential danger of deposits and stored or absorbed pollutants. The main target of all measures is to avoid any chemical

Characteristics of U-tube assembly design for CANDU 6 type steam generators

International Nuclear Information System (INIS)

Park, Jun Su; Jeong, Seung Ha

1996-06-01

Since the first operation of nuclear steam generator early 1960s, its performance requirements have been met but the steam generator problems have been met but the steam generator problems have been major cause of reducing the operational reliability, plant safety and availability. U-tube assembly of steam generator forms the primary system pressure boundary of the plant and have experienced several types of tube degradation problems. Tube failure and leakage resulting from the degradation will cause radioactive contamination of secondary system by the primary coolant, and this may lead to unplanned plant outages and costly repair operations such as tube plugging or steam generator replacement. For the case of steam generators for heavy water reactors, e.g. Wolsong 2, 3, and 4 NPP, a high cost of heavy water will be imposed additionally. During the plant operation, steam generator tubes can potentially be subject to adverse environmental conditions which will cause damages to U-tube assembly. Types of the damage depend upon the combined effects of design factors, materials and chemical environment of steam generator, and they are the pure water stress corrosion cracking, intergranular attack, pitting, wastage, denting, fretting and fatigue, etc. In this report, a comprehensive review of major design factors of recirculating steam generators has been performed against the potential tube damages. Then the design characteristics of CANDU-type Wolsong steam generator were investigated in detail, including tube material, thermalhydraulic aspects, tube-to-tubesheet joint, tube supports, water chemistry and sludge management. 9 tabs., 18 figs., 38 refs. (Author) .new

Numerical simulation of thermohydraulic behavior of the steam generator of PWR type reactor

International Nuclear Information System (INIS)

Braga, C.V.M.; Carajilescov, P.

1981-01-01

Generally, 'U' tube steam generators with natural internal recirculation are used in PWR power stations. A thermalhydraulic model is developed for simulation of such components, in steady state. The flow of the secondary cycle fluid is divided in two parts individually homogeneous, allowing for heat and mass exchange between them. The secondary pressure is determined by defining the moisture of the vapor that feeds the turbine. This model is applied to the Angra II steam generator, operating in nominal conditions and with tubing partially plugged. (Author) [pt

Attachment of iron corrosion products on steam generator tube and feed-water pump in PWRs secondary system

International Nuclear Information System (INIS)

Shoda, Y.; Ishihara, N.; Miyata, H.; Ohira, T.; Watanabe, Y.; Nonaka, Y.

2010-01-01

Operating experience of the secondary systems in PWRs indicates that scale attachment distinctly have an effect on the performance of water-steam cycle. Attached scale on outer surface of steam generator (SG) tube could induce many problems such as decrease heat efficiency of plant, corrosion of tube by intergranular attack (IGA), and choke of flow channel. Scale attached on rotor blade of feed water pump increases the driving steam consumption to keep the constant flow rate, and results in the thermal efficiency decrease of the plant. In this study, two types of test about scale deposition on equipment were executed in the conditions simulating the secondary system of PWR. One is SG model test, which simulated the circulating boiler composed of single SG tube and blow down line. The deposition rate under AVT condition was equivalent to plants revealed with extended period. High-AVT test provided useful reference, because the deposition rate of power plant is too small to measure in a short period after the beginning of High-AVT operation in Japan. The other is feed water pump model test. The mock-up pump is composed of a rotating stainless steel disk. As a result, it is confirmed that the deposition rate depends mostly on iron concentration in water and the exfoliation rate depends mainly on pH. Applying this information, the scale deposition-growth behavior on the equipment is quantitatively expressed by the model combined of scale deposition behavior and exfoliation behavior couples with the former. These results bring effective estimation for suppressing deposition-growth by the selection of water chemistry management and/or equipment improvement in the PWR secondary system. (author)

Steam generator group project progress report. Task 3, health physics

International Nuclear Information System (INIS)

Reece, W.D.; Hoenes, G.R.; Parkhurst, M.A.

1984-01-01

The gamma radiation fields in and around the retired Surry steam generator were measured extensively with thermoluminescent dosimeters (TLD's) and other standard radiation instruments. The techniques of measurement and the results are described for locations outside the shell, inside the channel head, and inside the secondary side of the steam generator. The gamma fields ranged from more than 10 R/hr in the middle of the tube bundle on the secondary side to less than 5 mR/hr at the bottom of the outside of the shell below the channel head. Co-60 was the only detected gamma emitter. The results of the measurements were used in an analytical model which predicted the Co-60 inventory to be between 70 and 87 curies

Steady-state heat transfer in an inverted U-tube steam generator

International Nuclear Information System (INIS)

Boucher, T.J.

1987-01-01

Experimental results are presented involving U-tube steam generator tube bundle local heat transfer and fluid conditions during stead-state, full-power operations performed at high temperatures and pressures with conditions typical of a pressurized water reactor (15.0 MPa primary pressure, 600 K steam generator inlet plenum fluid temperatures, 6.2 MPa secondary pressure). The Semiscale (MOD-2C facility represents the state-of-the-art in measurement of tube local heat transfer data and average tube bundle secondary fluid density at several elevations, which allows an estimate of the axial heat transfer and void distributions during steady-state and transient operations. The method of heat transfer data reduction is presented and the heat flux, secondary convective heat transfer coefficient, and void fraction distributions are quantified for steady-state, full-power operations

Hybrid preheat/recirculating steam generator

International Nuclear Information System (INIS)

Lilly, G.P.

1985-01-01

The patent describes a hybrid preheat/recirculating steam generator for nuclear power plants. The steam generator utilizes recirculated liquid to preheat incoming liquid. In addition, the steam generator incorporates a divider so as to limit the amount of recirculating water mixed with the feedwater. (U.K.)

Tube tightness survey during Phenix steam generator operation

International Nuclear Information System (INIS)

Cambillard, E.

1976-01-01

Phenix steam generators are once-through vessels with single-wall heat-exchange tubes. This design means that any leakage of water into the sodium must be detected as quickly as possible so that the installation can be shut down before extensive damage occurs. The detection of water leaks in Phenix steam generators is based on measurement of the concentration in the sodium, of hydrogen produced by the sodium-water reaction. Since the various modules--evaporators, superheaters, and reheaters--have no free sodium surfaces, detection of hydrogen in argon is not used in Phenix steam generators. The measurement systems employ a probe made of nickel tubes 0.3 mm thick. Hydrogen in the sodium diffuses into a chamber kept under vacuum by an ion pump. The hydrogen pressure in the chamber is measured by a quadrupole mass spectrometer. The nine measurement systems (three per steam generator) are calibrated by injecting hydrogen into the sodium of the secondary circuits. The data-processing computer calculates the hydrogen concentration in the sodium from the spectrometer signals and the probe temperatures, which are not regulated in Phenix; it generates instructions that enable the operator to act if a leak appears. So far, no leaks have been detected. These systems also make it possible to determine rates of hydrogen diffusion caused by corrosion of the steel walls on the water side

Design of a steam generator for PWR power plants and steady state simulation

International Nuclear Information System (INIS)

Ferreira, W.J.

1982-01-01

A procedure and a computer code for the thermal design of a steam generator for PWR power plants is developed. A vertical integral steam generator with inverted U-tubes and natural circulation of the secondary side is selected for modelling. Primary fluid velocity and recirculation ratio are varied to obtain the preliminary dimensions. Further, adjustments are made through iteractive solution of the equations of conservation of mass, energy and momentum. An agreement is found between design calculations for steam generators of different capacities and existing designs. (Author) [pt

Thermo-hydraulic stability study of a steam generator

International Nuclear Information System (INIS)

Magni, M C; Marcel, C P; Delmastro, D F

2012-01-01

In this work a mathematical model developed to investigate the thermalhydraulic stability of a helically coiled steam generator is presented. Such a steam generator is prone to experiment density wave oscillations. The model is therefore used to analyze the stability of the CAREM-25 reactor steam generators. The model is linear, numerically non-diffusive and nodal. In addition, it is able to represent non-uniform heat transfer fluxes between the primary and secondary coolant circuits. By using this model the marginal stability condition is found by varying the inlet friction coefficient for different conditions. The results are then compared with those obtained with a different model for which a simple uniform heat flux profiled is assumed. It is found that with such simplification the density waves instability mechanism is overestimated in a wide range of operating powers. For very low powers, in the contrary, the so-called uniform model underestimates the stabilizing inlet friction and therefore it gives non-conservative results. With the use of the more realistic non-uniform power profile model, it was possible to determine that, for a CAREM-25 steam generator, the most stable conditions is found at 60MW when the reactor operates at nominal pressure. Moreover, it is found that at high power levels the stability performance is dominated by the two-phase friction component while at low power levels the friction component originated in the over heated steam region prevail (author)

Design and manufacture of steam generators for replacement

International Nuclear Information System (INIS)

Hirano, Hiroshi; Kuri, Syuhei

1995-01-01

The basic specification of the steam generators for replacement as heat exchangers (the pressure, temperature, flow rate and thermal output on primary and secondary sides) is set same as that of steam generators before replacement, but the latest design reflecting the operation experience obtained so far and taking the countermeasures for preventing heating tube damage in it is adopted, such as the heating tubes made of TT 690 alloy, the tube support plates with four-lobe shape tube holes made of stainless steel, the stainless steel rest fittings of three in one set and so on. After the heating tube break accident in Mihama No. 2 plant, the quality control was further strengthened. The comparison of the specifications of the steam generators of respective plants before and after the replacement is shown. The main objective of improving steam generators is the heightening of the reliability of heating tubes against intergranular attack and primary water stress corrosion cracking. The improvements of heating tube material, tube support plate material, secondary side heat flow, the shape of tube holes of tube support plates, the method of expanding heating tubes, and vibration-controlling fittings are explained. As to the manufacturing procedure and quality control, the manufacture of raw materials, the build-up welding of tube plates, the manufacture of lower half shell plates, the tube hole making of support plates, the insection of outer cylinder, flow rate distribution plate. Support plates and heating tubes, the sealing welding and expanding of heating tubes, the fixing of rest fittings, the manufacture and fixing of water chamber cover, the manufacture of upper half shell, the fixing of parts inside it, the final joint and inspection are described. (K.I.)

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)

Advances in steam generator service technology

International Nuclear Information System (INIS)

Perez, Ric

1998-01-01

The most recent advances in pressurized water reactor steam generator service technology are discussed in this article. Focus is on new developments in robotics, including the Remotely Operated Service Arm (ROSA III); repair and maintenance services on the SG secondary side; and the newest advances in SG inspection. These products and services save utility costs, shorten outage durations, enhance plant performance and safety, and reduce radiation exposure. (author)

The steam generator programme of PISC III

International Nuclear Information System (INIS)

Birac, C.; Herkenrath, H.

1990-12-01

The PISC III Actions are intended to extend the results and methodologies of the previous PISC excercises, i.e. the validation of the capabilities of the various examination techniques when used on real defects in real components under real conditions of inspection. Being aware of the important safety role that steam generator tubes play as barrier between primary and secondary cooling system and of the industrial problems that the degradation of these tubes can create, the PISC III Management Board agreed to include in the PISC III Programme a special Action on Steam Generator Tubes Testing (SGT). It was decided to organize the programme in three phases, including Round Robin Tests (RRT): - capability tests on loose tubes, - capability tests on transportable mock-ups, - reliability tests on fixed mock-ups including some interesting SURRY tubes

Fluidelastic instability analysis of steam generator U-tubes at antivibration bar-inactive modes

International Nuclear Information System (INIS)

Lee, S.K.; Jo, J.C.

1995-01-01

This paper presents the results of thermal-hydraulic and fluidelastic U-tube instability analyses performed for the vertical type pressurized water reactor (PWR) steam generator model being employed at Kori units 2, 3 and 4, and Yonggwang units 1 and 2 in Korea. The thermal-hydraulic analysis for providing the detailed three-dimensional two-phase flow field in the secondary side of the steam generator was accomplished using the ATHOS3 steam generator thermal-hydraulic analysis code. The UTVA2 code designed for calculating both the free vibration responses and fluidelastic stability ratio of a specific U-tube under consideration was used to assess the potential for fluidelastic instability of the steam generator U-tubes at various conditions of antivibration bar (AVB)-inactive modes. The results of the fluidelastic instability analysis were discussed in comparison with those obtained for the steam generator U-tubes at AVB-active mode

Thermal circuit and supercritical steam generator of the BGR-300 nuclear power plant

International Nuclear Information System (INIS)

Afanas'ev, B.P.; Godik, I.B.; Komarov, N.F.; Kurochnkin, Yu.P.

1979-01-01

Secondary coolant circuit and a steam generator for supercritical steam parameters of the BGR-300 reactor plant are described. The BGR-300 plant with a 300 MW(e) high-temperature gas-cooled fast reactor is developed as a pilot commercial plant. It is shown that the use of a supercritical pressure steam increases the thermal efficiency of the plant and descreases thermal releases to the environment, permits to use home-made commercial turbine plants of large unit power. The proposed supercritical pressure steam generator has considerable advantages from the viewpoint of heat transfer and hydrodynamical processes

Steam generator water lancing

International Nuclear Information System (INIS)

Kamler, F.; Schneider, W.

1992-01-01

The tubesheet and tube support plate deposits in CANDU steam generators are notable for their hardness. Also notable is the wide variety of steam generator access situations. Because of the sludge hardness and the difficulty of the access, traditional water lancing processes which directed jets from the central tube free lane or from the periphery of the bundle have proven unsuitable. This has led to the need for some very unique inter tube water lancing devices which could direct powerful water jets directly onto the deposits. This type of process was applied to the upper broached plates of the Bruce A steam generators, which had become severely blocked. It has since been applied to various other steam generator situations. This paper describes the flexlance equipment development, qualification, and performance in the various CANDU applications. 4 refs., 2 tabs., 7 figs

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

Modelling of WWER-1000 steam generators by REALP5/MOD3.2 code

Energy Technology Data Exchange (ETDEWEB)

D`Auria, F.; Galassi, G.M. [Univ. of Pisa (Italy); Frogheri, M. [Univ. of Genova (Italy)

1997-12-31

The presentation summarises the results of best estimate calculations carried out with reference to the WWER-1000 Nuclear Power Plant, utilizing a qualified nodalization set-up for the Relap5/Mod3.2 code. The nodalization development has been based on the data of the Kozloduy Bulgarian Plant. The geometry of the steam generator imposed drastic changes in noding philosophy with respect to what is suitable for the U-tubes steam generators. For the secondary side a symmetry axis was chosen to separate (in the nodalization) the hot and the cold sides of the tubes. In this way the secondary side of the steam generators was divided into three zones: (a) the hot zone including the hot collector and the hot l/2 parts of the tubes; (b) the cold zone including the cold collector and the cold 1/2 parts of the tubes; (c) the downcomer region, where down flow is assumed. As a consequence of above in the primary side more nodes are placed on the hot side of the tubes. Steady state and transient qualification has been achieved, considering the criteria proposed at the University of Pisa, utilizing plant transient data from the Kozloduy and the Ukrainian Zaporosche Plants. The results of the application of the qualified WWER-1000 Relap5/Mod3.2 nodalization to various transients including large break LOCA, small break LOCA and steam generator tube rupture, together with a sensitivity analysis on the steam generators, are reported in the presentation. Emphasis is given to the prediction of the steam generators performances. 23 refs.

Modelling of WWER-1000 steam generators by REALP5/MOD3.2 code

Energy Technology Data Exchange (ETDEWEB)

D` Auria, F; Galassi, G M [Univ. of Pisa (Italy); Frogheri, M [Univ. of Genova (Italy)

1998-12-31

The presentation summarises the results of best estimate calculations carried out with reference to the WWER-1000 Nuclear Power Plant, utilizing a qualified nodalization set-up for the Relap5/Mod3.2 code. The nodalization development has been based on the data of the Kozloduy Bulgarian Plant. The geometry of the steam generator imposed drastic changes in noding philosophy with respect to what is suitable for the U-tubes steam generators. For the secondary side a symmetry axis was chosen to separate (in the nodalization) the hot and the cold sides of the tubes. In this way the secondary side of the steam generators was divided into three zones: (a) the hot zone including the hot collector and the hot l/2 parts of the tubes; (b) the cold zone including the cold collector and the cold 1/2 parts of the tubes; (c) the downcomer region, where down flow is assumed. As a consequence of above in the primary side more nodes are placed on the hot side of the tubes. Steady state and transient qualification has been achieved, considering the criteria proposed at the University of Pisa, utilizing plant transient data from the Kozloduy and the Ukrainian Zaporosche Plants. The results of the application of the qualified WWER-1000 Relap5/Mod3.2 nodalization to various transients including large break LOCA, small break LOCA and steam generator tube rupture, together with a sensitivity analysis on the steam generators, are reported in the presentation. Emphasis is given to the prediction of the steam generators performances. 23 refs.

Steam generator tube integrity program

International Nuclear Information System (INIS)

Dierks, D.R.; Shack, W.J.; Muscara, J.

1996-01-01

A new research program on steam generator tubing degradation is being sponsored by the U.S. Nuclear Regulatory Commission (NRC) at Argonne National Laboratory. This program is intended to support a performance-based steam generator tube integrity rule. Critical areas addressed by the program include evaluation of the processes used for the in-service inspection of steam generator tubes and recommendations for improving the reliability and accuracy of inspections; validation and improvement of correlations for evaluating integrity and leakage of degraded steam generator tubes, and validation and improvement of correlations and models for predicting degradation in steam generator tubes as aging occurs. The studies will focus on mill-annealed Alloy 600 tubing, however, tests will also be performed on replacement materials such as thermally-treated Alloy 600 or 690. An overview of the technical work planned for the program is given

Steam generator tube extraction

International Nuclear Information System (INIS)

Delorme, H.

1985-05-01

To enable tube examination on steam generators in service, Framatome has now developed a process for removing sections of steam generator tubes. Tube sections can be removed without being damaged for treating the tube section expanded in the tube sheet

International examples of steam generator replacement

International Nuclear Information System (INIS)

Wiechmann, K.

1993-01-01

Since 1979-1980 a total of twelve nuclear power plants world-wide have had their steam generators replaced. The replacement of the Combustion steam generators in the Millstone-2 plant in the United States was completed very recently. Steam generator replacement activities are going on at present in four plants. In North Anna, the steam generators have been under replacement since January 1990. In Japan, preparations have been started for Genkai-1. Since January 1992, the two projects in Beznau-1, Switzerland, and Doel-3, Belgium, have bee planned and executed in parallel. Why steam generator replacement? There are a number of defect mechanisms which give rise to the need for early steam generator replacement. One of the main reasons is the use of Inconel-600 as material for the heating tubes. Steam generator heating tubes made of Inconel-600 have been known to exhibit their first defects due to stress corrosion cracking after less than one year of operation. (orig.) [de

Cheaper power generation from surplus steam generating capacities

International Nuclear Information System (INIS)

Gupta, K.

1996-01-01

Prior to independence most industries had their own captive power generation. Steam was generated in own medium/low pressure boilers and passed through extraction condensing turbines for power generation. Extraction steam was used for process. With cheaper power made available in Nehru era by undertaking large hydro power schemes, captive power generation in industries was almost abandoned except in sugar and large paper factories, which were high consumers of steam. (author)

WWER steam generator tube structural and leakage integrity

International Nuclear Information System (INIS)

Splichal, K.; Krhounek, Vl.; Otruba, J.; Ruscak, M.

1998-01-01

The integrity of heat exchange tubes may influence the lifetime of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirements are to assure very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evaluation and heat exchange tubes plugging. The stress corrosion cracking and pitting are the main corrosion damages of WWER heat exchange tubes and are initiated from the outer surface. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through wall cracks, oriented preferentially in the axial direction. The paper presents the leakage and plugging limits for WWER steam generators, which have been determined from leak tests and burst tests. The tubes with axial part-through and through-wall defects have been used. The permissible value of primary to secondary leak rate was evaluated with respect to permissible axial through-wall defect size of WWER 440 and 1000 steam generator tubes. Blocking of the tube cracks by corrosion product particles and other compounds reduces the primary to secondary leak rate. The plugging limits involve the following factors: permissible tube wall thickness which determine further operation of the tubes with defects and assures their integrity under operating conditions and permissible size of a through-wall crack which is sufficiently stable under normal and accident conditions in relation to the critical crack length. For the evaluation of burst test of heat exchange tubes with longitudinal through-wall defects the instability criterion has been used and the dependence of the normalised burst pressure on the normalised length of an axial through-wall defect has been determined. The validity of the criterion of instability for WWER tubes with through

Review of EPRI's steam generator R and D program

International Nuclear Information System (INIS)

Millett, P.J.; Welty, C.J.

1998-01-01

EPRI has carried out an extensive R and D program on SG technology since the mid 1970's. Very early efforts under the auspices of the Steam Generator Owners Group (SGOG) focused on developing remedial actions for the critical SG corrosion issues of denting, wastage and pitting. Fundamental work was also carried out in the development of thermal hydraulic models for vibration and wear, chemical cleaning and tube repair techniques. In the late 1980's and continuing through today, the program has shifted emphasis towards management of steam generator degradation, primarily stress corrosion cracking of the SG tubes on both the primary and secondary sides. The current Steam Generator Management Program (SGMP) carries out R and D in four areas; materials, chemistry, thermal hydraulics and non-destructive testing. The strategic goals of this program and projects put in place to achieve these goals will be reviewed in detail in this paper. (author)

Fail-safety of the EBR-II steam generator system

International Nuclear Information System (INIS)

Chopra, P.S.; Stone, C.C.; Hutter, E.; Barney, W.K.; Staker, R.G.

1976-01-01

Fail-safe analyses of the EBR-II steam-generator system show that a postulated non-instantaneous leak of water or steam into sodium, through a duplex tube or a tubesheet, at credible leak rates will not structurally damage the evaporators and superheaters. However, contamination of the system and possible shell wastage by sodium-water reaction products may render the system inoperable for a period exceeding six months. This period would be shortened to three months if the system were modified by adding a remotely operated water dump system, a steam vent system, a secondary sodium superheater relief line, and a tubesheet leak-detection system

Degradation monitoring in IRIS steam generators

International Nuclear Information System (INIS)

Alpay, B.; Holloway, J. P.; Lee, J. C.

2007-01-01

We present a degradation monitoring technique based on unscented Kalman filtering (UKF), which uses a nonlinear system model without linearization to estimate the status of the component/state variables. To test the applicability of the methodology, the fouling of tubes is chosen among various degradation mechanisms for the IRIS (International Reactor Innovative and Secure) steam generators (SGs). The degradation monitoring algorithm diagnoses the tube fouling and estimates the thickness of the crud deposited on the secondary side of the SG along with the increase in the pressure drop triggered by fouling. A stand-alone SG model developed with the RELAP5 code was used to simulate the transient behavior of the SG and drive an UKF state estimate. By using the secondary side outlet temperature as the measurement and the nodal pressures along the secondary side as states, UKF generated accurate estimates of the crud layer thicknesses for different crud formations. (authors)

Evaluation of sulfur hexafluoride and helium for steam generator leak location: Final report

International Nuclear Information System (INIS)

Kassen, W.R.

1987-01-01

Since the use of sulfur hexafluoride as a tracer for identifying sources of primary to secondary leakage in PWR steam generators appeared to offer significant sensitivity advantages, the thermal stability of sulfur hexafluoride in water was evaluated at steam generator operating temperature. Significant decomposition was observed after 2 to 4 hours at temperature. Key decomposition products were fluoride and sulfide ions. Based on this observation and these limited test results, the use of SF 6 for PWR steam generator leak location can not be recommended at this time. A survey of 15 utilities was conducted in regard to their application experience with the helium tracer-mass spectroscopy technique for steam generator leak location. Although several successful steam generator integrity programs do not include use of this technique, it has proven to be a useful addition to the inspection program at some plants. No corrosion concerns appear to be associated with this technique

Steam generator for nuclear reactors

International Nuclear Information System (INIS)

Byerley, W.M.; Bennett, R.R.

1978-01-01

In the steam generator, the primary medium is led through a U-shaped tube bundle heating up a secondary medium (feedwater) which flows around the tube bundle via a preheating chamber. In order to optimize heat transfer inside the preheating chamber, the feedwater is separated into a counterflow and a parallel flow with regard to the primary medium by means of partitioning walls and deflectors. The ratio is 70/30%. This way, boiling in the preheater is avoided, i.e. the high LMTD (logaritmic mean temperature difference) is fully utilized. (DG) [de

Diagnostic system of steam generator, especially molten metal heated steam generator

International Nuclear Information System (INIS)

Matal, O.; Martoch, J.

1986-01-01

A diagnostic system is described and graphically represented consisting of a leak detector, a medium analyzer and sensors placed on the piping connected to the indication sections of both tube plates. The advantage of the designed system consists in the possibility of detecting tube failure immediately on leak formation, especially in generators with duplex tubes. This shortens the period of steam generator shutdown for repair and reduces power losses. The design also allows to make periodical leak tests during planned steam generator shutdowns. (A.K.)

Mitigation of caustic stress corrosion cracking of steam generator tube materials by blowdown -a case study

International Nuclear Information System (INIS)

Dutta, Anu; Patwegar, I.A.; Chaki, S.K.; Venkat Raj, V.

2000-01-01

The vertical U-tube steam generators are among the most important equipment in nuclear power plants as they form the vital link between the reactor and the turbogenerator. Over ∼ 35 years of operating experience of water cooled reactor has demonstrated that steam generator tubes are susceptible to various forms of degradation. This degradation leads to failure and outages of the power plant. A majority of these failures have been attributed to concentrated alkali attacks in the low flow areas such as crevices in the tube to tube sheet joints, baffle plate location and the areas of sludge deposits. Free hydroxides can be produced by improper maintenance of phosphate chemical control in the secondary side of the steam generators and also by the thermal decomposition of impurities present in the condenser cooling water which may leak into the feed water through the condenser tubes. The free hydroxides concentrate in the low flow areas. This buildup of free hydroxide in combination with residual stress leads to caustic stress corrosion cracking. In order to mitigate caustic stress corrosion cracking of Inconel 600 tubes, the trend is to avoid phosphate dosing. Instead All Volatile Treatment (AVT) for secondary water is used backed by full flow condensate polishing. Sodium hydroxide concentration is now being considered as the basis for steam generator blowdown. A methodology has been established for determining the blowdown requirement in order to mitigate caustic stress corrosion cracking in the secondary side of the vertical U-tube natural circulation steam generator. A case study has been carried out for zero solid treatment (AVT coupled with full flow condensate polishing plant) water chemistry. Only continuous blowdown schemes have been studied based on maximum caustic concentration permissible in the secondary side of the steam generator. The methodology established can also be used for deciding concentration of any other impurities

Structural and leakage integrity assessment of WWER steam generator tubes

International Nuclear Information System (INIS)

Splichal, K.; Otruba, J.; Keilova, E.; Krhounek, V.; Turek, J.

1996-01-01

The leakage and plugging limits were derived for WWER steam generators based on leak and burst tests using tubes with axial part-through and through-wall defects. The following conclusions were arrived at: (i) The permissible primary-to-secondary leak rate with respect to the permissible through-wall defect size of WWER-440 and WWER-1000 steam generator tubes is 8 l/h. (ii) The primary-to-secondary leak rate is reduced by the blocking of the tube cracks by corrosion product particles and other substances. (iii) The rate of crack penetration through the tube wall is higher than the crack widening. (iv) The validity of the criterion of instability for tubes with through-wall cracks was confirmed experimentally. For the WWER-440 and WWER-1000 steam generators, the critical size of axial through-wall cracks, for the threshold primary-to-secondary pressure difference, is 13.8 and 12.0 mm, respectively. (v) The calculated leakage for the rupture of one tube and for the assumed extreme defects is two orders and one order of magnitude, respectively, higher than the proposed primary water leakage limit of 8 l/h. (vi) The experiments gave evidence that the use of the permissible thinning limit of 80% for the heat exchange tube plugging does not bring about uncontrollable leakage or unstable crack growth. This is consistent with experience gained at WWER-440 type nuclear power plants. 4 tabs., 5 figs., 9 refs

Steam generator with perfected dryers

International Nuclear Information System (INIS)

Fenet, J.C.

1987-01-01

This steam generator has vertically superposed array of steam dryers. These dryers return the steam flow of 180 0 . The return of the water is made by draining channels to the steam production zone [fr

ANL/CANTIA code for steam generator tube integrity assessment

International Nuclear Information System (INIS)

Revankar, S.T.; Wolf, B.; Majumdar, S.; Riznic, J.R.

2009-01-01

Steam generator (SG) tubes have an important safety role in CANDU type reactors and Pressurized Water Reactors (PWR) because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear plant. The SG tubes are susceptible to corrosion and damage. A failure of a single steam generator tube, or even a few tubes, would not be a serious safety-related event in a CANDU reactor. The leakage from a ruptured tube is within makeup capacity of the primary heat transport system, so that as long as the operator takes the correct actions, the off-site consequences will be negligible. A sufficient safety margin against tube rupture used to be the basis for a variety of maintenance strategies developed to maintain a suitable level of plant safety and reliability. Several through-wall flaws may remain in operation and potentially contribute to the total primary-to-secondary leak rate. Assessment of the conditional probabilities of tube failures, leak rates, and ultimately risk of exceeding licensing dose limits has been used for steam generator tube fitness-for-service assessment. The advantage of this type of analysis is that it avoids the excessive conservatism typically present in deterministic methodologies. However, it requires considerable effort and expense to develop all of the failure, leakage, probability of detection, and flaw growth distributions and models necessary to obtain meaningful results from a probabilistic model. The Canadian Nuclear Safety Commission (CNSC) recently developed the CANTIA methodology for probabilistic assessment of inspection strategies for steam generator tubes as a direct effect on the probability of tube failure and primary-to-secondary leak rate Recently Argonne National Laboratory has developed tube integrity and leak rate models under Integrated Steam Generator Tube Integrity Program (ISGTIP-2). These models have been incorporated in the ANL/CANTIA code. This paper presents the ANL

Effects of Chemistry Parameters of Primary Water affecting Leakage of Steam Generator Tube Cracks

Energy Technology Data Exchange (ETDEWEB)

Shin, D. M.; Cho, N. C.; Kang, Y. S.; Lee, K. H. [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

Degradation of steam generator (SG) tubes can affect pressure boundary tightness. As a defense-in-depth measure, primary to secondary leak monitoring program for steam generators is implemented, and operation is allowed under leakage limits in nuclear power plants. Chemistry parameters that affect steam generator tube leakage due to primary water stress corrosion cracking (PWSCC) are investigated in this study. Tube sleeves were installed to inhibit leakage and improve tube integrity as a part of maintenance methods. Steam generators occurred small leak during operation have been replaced with new steam generators according to plant maintenance strategies. The correlations between steam generator leakage and chemistry parameters are presented. Effects of primary water chemistry parameters on leakage from tube cracks were investigated for the steam generators experiencing small leak. Unit A experienced small leakage from steam generator tubes in the end of operation cycle. It was concluded that increased solubility of oxides due to high pHT could make leakage paths, and low boron concentration lead to less blockage in cracks. Increased dissolved hydrogen may retard crack propagations, but it did not reduce leak rate of the leaking steam generator. In order to inhibit and reduce leakage, pH{sub T} was controlled by servicing cation bed operation. The test results of decreasing pHT indicate low pHT can reduce leak rate of PWSCC cracks in the end of cycle.

Effects of Chemistry Parameters of Primary Water affecting Leakage of Steam Generator Tube Cracks

International Nuclear Information System (INIS)

Shin, D. M.; Cho, N. C.; Kang, Y. S.; Lee, K. H.

2016-01-01

Degradation of steam generator (SG) tubes can affect pressure boundary tightness. As a defense-in-depth measure, primary to secondary leak monitoring program for steam generators is implemented, and operation is allowed under leakage limits in nuclear power plants. Chemistry parameters that affect steam generator tube leakage due to primary water stress corrosion cracking (PWSCC) are investigated in this study. Tube sleeves were installed to inhibit leakage and improve tube integrity as a part of maintenance methods. Steam generators occurred small leak during operation have been replaced with new steam generators according to plant maintenance strategies. The correlations between steam generator leakage and chemistry parameters are presented. Effects of primary water chemistry parameters on leakage from tube cracks were investigated for the steam generators experiencing small leak. Unit A experienced small leakage from steam generator tubes in the end of operation cycle. It was concluded that increased solubility of oxides due to high pHT could make leakage paths, and low boron concentration lead to less blockage in cracks. Increased dissolved hydrogen may retard crack propagations, but it did not reduce leak rate of the leaking steam generator. In order to inhibit and reduce leakage, pH_T was controlled by servicing cation bed operation. The test results of decreasing pHT indicate low pHT can reduce leak rate of PWSCC cracks in the end of cycle

Analysis of density wave instability in counter-flow steam generators using STEAMFREQ-X

International Nuclear Information System (INIS)

Chan, K.C.; Yadigaroglu, G.

1986-01-01

The STEAMFREQ-X computer model was developed to provide a more comprehensive modeling of the different phenomena that are important to stability analysis of counter-flow steam generators. It uses a frequency-domain analysis and considers heat-flux/flow coupling between the primary and secondary fluids in space and time. Predictions by STEAMFREQ-X were compared with data from both a multi-channel liquid-sodium heated steam generator and a set of single pipe test data. Predicted outlet steam qualities at instability thresholds were within 15% of experimental data for all test points. (orig.)

Steam generator reliability improvement project

International Nuclear Information System (INIS)

Blomgren, J.C.; Green, S.J.

1987-01-01

Upon successful completion of its research and development technology transfer program, the Electric Power Research Institute's Steam Generator Owners Group (SGOG II) will disband in December 1986 and be replaced in January 1987 by a successor project, the Steam Generator Reliability Project (SGRP). The new project, funded in the EPRI base program, will continue the emphasis on reliability and life extension that was carried forward by SGOG II. The objectives of SGOG II have been met. Causes and remedies have been identified for tubing corrosion problems, such as stress corrosion cracking and pitting, and steam generator technology has been improved in areas such as tube wear prediction and nondestructive evaluation (NDE). These actions have led to improved reliability of steam generators. Now the owners want to continue with a centrally managed program that builds on what has been learned. The goal is to continue to improve steam generator reliability and solve small problems before they become large problems

Steam generator reliability improvement project

International Nuclear Information System (INIS)

Blomgren, J.C.; Green, S.J.

1987-01-01

Upon successful completion of its research and development technology transfer program, the Electric Power Research Institute's (EPRI's) Steam Generator Owners Group (SGOG II) will disband in December 1986, and be replaced in January 1987, by a successor project, the Steam Generator Reliability Project (SGRP). The new project, funded in the EPRI base program, will continue to emphasize reliability and life extension, which were carried forward by SGOG II. The objectives of SGOG II have been met. Causes and remedies have been identified for tubing corrosion problems such as stress corrosion cracking and pitting, and steam generator technology has been improved in areas such as tube wear prediction and nondestructive evaluation. These actions have led to improved reliability of steam generators. Now the owners want to continue with a centrally managed program that builds on what has been learned. The goal is to continue to improve steam generator reliability and to solve small problems before they become large problems

How safe is defect specific maintenance of steam generator tubes?

International Nuclear Information System (INIS)

Dvorsek, T.; Cizelj, L.

1995-01-01

Outside diameter stress corrosion cracking at the tube to tube support plate intersections is assessed in the paper. The impact of defect specific maintenance on steam generator operation safety and reliability was investigated. This was performed by comparing efficiencies of defect specific and traditional maintenance strategy. The efficiency was studied through expected primary-to-secondary leak rate and tube rupture probability in a case of postulated accidental operating conditions, and number of tubes which shall be plugged using both maintenance strategies. In general, the efficiency of specific maintenance is function of particular steam generator and operating cycle. (author)

The effect of steam separataor efficiency on transient following a steam line break

International Nuclear Information System (INIS)

Choi, J.H.; Ohn, M.Y.; Lee, N.H.; Hwang, S.T.; Lee, S.K.

1996-01-01

Detailed thermalhydraulic simulations for CANDU 6 steam line break inside containment are performed to predict the response of the primary and secondary circuits. The analysis is performed using the thermalhydraulic computer code, CATHENA, with a coupled primary and secondary circuit model. A two-loop representation of the primary and secondary circuits is modelled. The secondary circuit model includes the feedwater line from the deaerator storage tank, multi-node steam generators and the steam line up to the turbine. Two cases were carried out using different assumptions for the efficiency of the steam separators. Case 1 assumes the efficiency of the steam separators becomes zero when the water level in the steam drum increases to the elevation of primary cyclones, or the outlet flow from the steam generator becomes higher than 150 % of normal flow. Case 2 assumes the efficiency becomes zero only when the water level in the steam drum reaches the elevation of primary cyclones. The simulation results show that system responses are sensitive to the assumption for the efficiency of the steam separators and case 1 gives higher discharge energy. Fuel cooling is assured, since primary circuit is cooled down sufficiently by the steam generators for both cases. (author)

PWR steam generator chemical cleaning. Phase I: solvent and process development. Volume II

International Nuclear Information System (INIS)

Larrick, A.P.; Paasch, R.A.; Hall, T.M.; Schneidmiller, D.

1979-01-01

A program to demonstrate chemical cleaning methods for removing magnetite corrosion products from the annuli between steam generator tubes and the tube support plates in vertical U-tube steam generators is described. These corrosion products have caused steam generator tube ''denting'' and in some cases have caused tube failures and support plate cracking in several PWR generating plants. Laboratory studies were performed to develop a chemical cleaning solvent and application process for demonstration cleaning of the Indian Point Unit 2 steam generators. The chemical cleaning solvent and application process were successfully pilot-tested by cleaning the secondary side of one of the Indian Point Unit 1 steam generators. Although the Indian Point Unit 1 steam generators do not have a tube denting problem, the pilot test provided for testing of the solvent and process using much of the same equipment and facilities that would be used for the Indian Point Unit 2 demonstration cleaning. The chemical solvent selected for the pilot test was an inhibited 3% citric acid-3% ascorbic acid solution. The application process, injection into the steam generator through the boiler blowdown system and agitation by nitrogen sparging, was tested in a nuclear environment and with corrosion products formed during years of steam generator operation at power. The test demonstrated that the magnetite corrosion products in simulated tube-to-tube support plate annuli can be removed by chemical cleaning; that corrosion resulting from the cleaning is not excessive; and that steam generator cleaning can be accomplished with acceptable levels of radiation exposure to personnel

Dynamic performance investigation of once-through-type steam generator for NPP using a large-scale model

International Nuclear Information System (INIS)

Kats, F.M.; Ostrovskij, L.A.; Ehskin, N.B.

1985-01-01

An experimental bench is described as well as the results of dynamic performance investigation of mass- and heat transfer of the once-through type steam generator for the NPP with weak superheat. Coolant for the primary and secondary circuit is water. Under investigation conditions the possibility of changin.o primary and secondary circuit temperatures has been supphed as well as the primary circuit flow rate and the secondary circuit pressure changes. Transients for differen.t operating conditions are considered. The possibility for construction of the steam generator automatic control system is based

49 CFR 229.105 - Steam generator number.

Science.gov (United States)

2010-10-01

... 49 Transportation 4 2010-10-01 2010-10-01 false Steam generator number. 229.105 Section 229.105..., DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Safety Requirements Steam Generators § 229.105 Steam generator number. An identification number shall be marked on the steam generator's...

Aerosol retention in the flooded steam generator bundle during SGTR

International Nuclear Information System (INIS)

Lind, Terttaliisa; Dehbi, Abdel; Guentay, Salih

2011-01-01

Research highlights: → High retention of aerosol particles in a steam generator bundle flooded with water. → Increasing particle inertia, i.e., particle size and velocity, increases retention. → Much higher retention of aerosol particles in the steam generator bundle flooded with water than in a dry bundle. → Much higher retention of aerosol particles in the steam generator bundle than in a bare pool. → Bare pool models have to be adapted to be applicable for flooded bundles. - Abstract: A steam generator tube rupture in a pressurized water reactor may cause accidental release of radioactive particles into the environment. Its specific significance is in its potential to bypass the containment thereby providing a direct pathway of the radioactivity from the primary circuit to the environment. Under certain severe accident scenarios, the steam generator bundle may be flooded with water. In addition, some severe accident management procedures are designed to minimize the release of radioactivity into the environment by flooding the defective steam generator secondary side with water when the steam generator has dried out. To extend our understanding of the particle retention phenomena in the flooded steam generator bundle, tests were conducted in the ARTIST and ARTIST II programs to determine the effect of different parameters on particle retention. The effects of particle type (spherical or agglomerate), particle size, gas mass flow rate, and the break submergence on particle retention were investigated. Results can be summarized as follows: increasing particle inertia was found to increase retention in the flooded bundle. Particle shape, i.e., agglomerate or spherical structure, did not affect retention significantly. Even with a very low submergence, 0.3 m above the tube break, significant aerosol retention took place underlining the importance of the jet-bundle interactions close to the tube break. Droplets were entrained from the water surface with

LMFBR steam generator development: duplex bayonet tube steam generator. Volume II

International Nuclear Information System (INIS)

DeFur, D.D.

1975-04-01

This report represents the culmination of work performed in fulfillment of ERDA Contract AT(11-1)-2426, Task Agreement 2, in which alternate steam generator designs were developed and studied. The basic bayonet tube generator design previously developed by C-E under AEC Contract AT(11-1)-3031 was expanded by incorporating duplex heat transfer tubes to enhance the unit's overall safety and reliability. The effort consisted of providing and evaluating conceptual designs of the evaporator, superheater and reheater components for a large plant LMFBR steam generator (950 MWt per heat transport loop)

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)

Scale Thickness Measurement of Steam Generator Tubing Using Eddy Current Signal of Bobbin Coil

International Nuclear Information System (INIS)

Kim, Chang Soo; Um, Ki Soo; Kim, Jae Dong

2012-01-01

Steam generator is one of the major components of nuclear power plant and steam generator tubes are the pressure boundary between primary and secondary side, which makes them critical for nuclear safety. As the operating time of nuclear power plant increases, not only damage mechanisms but also scaled deposits on steam generator tubes are known to be problematic causing tube support flow hole blockage and heat fouling. The ability to assess the extent and location of scaled deposits on tubes became essential for management and maintenance of steam generator and eddy current bobbin data can be utilized to measure thickness of scale on tubes. In this paper, tube reference standards with various thickness of scaled deposit has been set up to provide information about the overall deposit condition of steam generator tubes, providing essential tool for steam generator management and maintenance to predict and prevent future damages. Also, methodology to automatically measure scale thickness on tubes has been developed and applied to field data to estimate overall scale amount.

SGTR Project: Separate Effect Studies for Vertical Steam Generators

Energy Technology Data Exchange (ETDEWEB)

Peyres, V.; Polo, J.; Herranz, L. E.

2003-07-01

The SGTR project has been carried out within the fifth EURATOM Framework Programme (Contract No FIKS-CT-1999-0007). Its main objective was to provide an experimental database and to develop and/or verify models to support definition of accident management measures in the hypothetical case of a Steam Generator tube Rupture (SGTR) sequence. The project addressed both vertical and horizontal steam generator designs. This report summarises the main results obtained in the intermediate scale experimentation that addressed Western type steam generators. The specific goal of this test programme was to investigate aerosol retention in the break stage of the secondary side of a water-empty steam generator. The test matrix consisted of 12 tests that explored the influence of variables such as break type and orientation and inlet gas flow rate. This work was performed in the PECA facility of the Laboratory for Analysis of Safety Systems (LASS). Aerosol retention at the break stage of a dry steam generator was observed to be low and non-uniform. Neither break type nor orientation affected results significantly whenever gas flowrates exceeded about 100 kg/h. However, deposition patterns guillotine breaks and fish mouth ones showed remarkable differences. For flowrates above 100 kg/hm the higher the gas flow velocity, the lower the total mass depleted on tube bundle surfaces; however, at lower flowrates this trend was not maintained. An attempt to measure gas injection velocity at the break exit by Particle Image Velocity (PIV) was done but data were highly uncertain. (Author) 2 refs.

Scale model test results for an inverted U-tube steam generator with comparisons to heat transfer correlations

International Nuclear Information System (INIS)

Boucher, T.J.

1987-01-01

To provide data for assessment and development of thermal-hydraulic computer codes, bottom main feedwater-line-break transient simulations were performed in a scale model (Semiscale Mod-2C) of a pressurized water reactor (PWR) with conditions typical of a PWR (15.0 MPa primary pressure, 600 K steam generator inlet plenum fluid temperatures, 6.2 MPa secondary pressure). The state-of-the-art measurements in the scale model (Type III) steam generator allow for the determination of U-tube steam generator allow for the determination of U-tube steam generator secondary component interactions, tube bundle local radial heat transfer, and tube bundle and riser vapor void fractions for steady state and transient operations. To enhance the understanding of the observed phenomena, the component interactions, local heat fluxes, local secondary convective heat transfer coefficients and local vapor void fractions are discussed for steady state, full-power and transient operations. Comparisons between the measurement-derived secondary convective heat transfer coefficients and those predicted by a number of correlations, including the Chen correlation currently used in thermal-hydraulic computer codes, show that none of the correlations adequately predict the data and points out the need for the formulation of a new correlation based on this experimental data. The unique information presented herein should be of the interest to anyone involved in modeling inverted U-tube steam generator thermal-hydraulics for forced convection boiling/vaporization heat transfer. 5 refs., 13 figs., 1 tab

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)

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

AGE RELATED DEGRADATION OF STEAM GENERATOR INTERNALS BASED ON INDUSTRY RESPONSES TO GENERIC LETTER 97-06

International Nuclear Information System (INIS)

SUBUDHI, M.; SULLIVAN, JR. E.J.

2002-01-01

THIS PAPER PRESENTS THE RESULTS OF AN AGING ASSESSMENT OF THE NUCLEAR POWER INDUSTRY RESPONSES TO NRC GENERIC LETTER 97-06 ON THE DEGRADATION OF STEAM GENERATOR INTERNALS EXPERIENCED AT ELECTRICITE DE FRANCE (EDF) PLANTS IN FRANCE AND AT A UNITED STATES PRESSURIZED WATER REACTOR (PWR). WESTINGHOUSE (W), COMBUSTION ENGINEERING (CE), AND BABCOCK AND WILCOX (BW) STEAM GENERATOR MODELS, CURRENTLY IN SERVICE AT U.S. NUCLEAR POWER PLANTS, POTENTIALLY COULD EXPERIENCE DEGRADATION SIMILAR TO THATFOUND AT EDF PLANTS AND THE U.S. PLANT. THE STEAM GENERATORS IN MANY OF THE U.S. PWRS HAVE BEEN REPLACED WITH STEAM GENERATORS WITH STEAM GENERATORS WITH IMPROVED DESIGNS AND MATERIALS. THESE REPLACEMENT STEAM GENERATORS HAVE BEEN MANUFACTURED IN THE U.S. AND ABROAD. DURING THIS ASSESSMENT, EACH OF THE THREE OWNERS GROUPS (W,CE, AND BW) IDENTIFIED FOR ITS STEAM GENERATOR, MODELS ALL THE POTENTIAL INTERNAL COMPONENTS THAT ARE VULNERABLE TO DEGRADATION WHILE IN SERVICE. EACH OWNERS GROUPDEVELOPED INSPEC TION AND MONITORING GUIDANCE AND RECOMMENDATIONS FOR ITS PARTICULAR STEAM GENERATOR MODELS. THE NUCLEAR ENERGY INSTITUTE INCORPORATED IN NEI 97-06 STEAM GENERATOR PROGRAM GUIDELINES, A REQUIREMENT TO MONITOR SECONDARY SIDE STEAM GENERATOR COMPONENTS IF THEIR FAILURE COULD PREVENT THE STEAM GENERATOR FROM FULFILLING ITS INTENDED SAFETY-RELATED FUNCTION. LICENSEES INDICATED THAT THEY IMPLEMENTED OR PLANNED TO IMPLEMENT, AS APPROPRIATE FOR THEIR STEAM GENERATORS, THEIR OWNERS GROUPRECOMMENDATIONS TO ADDRESS THE LONG-TERM EFFECTS OF THE POTENTIAL DEGRADATION MECHANISMS ASSOCIATED WITH THE STEAM GENERATOR INTERNALS

Key findings from the artist project on aerosol retention in a dry steam generator

International Nuclear Information System (INIS)

Dehbi, Abedeloahab; Suckow, Deltef; Lind, Tettaliisa; Guentat, Salih; Danner, Steffen; Mukin, Roman

2016-01-01

A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program

Key findings from the artist project on aerosol retention in a dry steam generator

Energy Technology Data Exchange (ETDEWEB)

Dehbi, Abedeloahab; Suckow, Deltef; Lind, Tettaliisa; Guentat, Salih; Danner, Steffen; Mukin, Roman [Nuclear Energy and Safety Research Department, Paul Scherrer Institute, Villigen (Switzerland)

2016-08-15

A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.

Key Findings from the Artist Project on Aerosol Retention in a Dry Steam Generator

Directory of Open Access Journals (Sweden)

Abdelouahab Dehbi

2016-08-01

Full Text Available A steam generator tube rupture (SGTR event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI initiated the Aerosol Trapping In Steam GeneraTor (ARTIST Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.

Options for Steam Generator Decommissioning

International Nuclear Information System (INIS)

Krause, Gregor; Amcoff, Bjoern; Robinson, Joe

2016-01-01

Selecting the best option for decommissioning steam generators is a key consideration in preparing for decommissioning PWR nuclear power plants. Steam Generators represent a discrete waste stream of large, complex items that can lend themselves to a variety of options for handling, treatment, recycling and disposal. Studsvik has significant experience in processing full size Steam Generators at its metal recycling facility in Sweden, and this paper will introduce the Studsvik steam generator treatment concept and the results achieved to date across a number of projects. The paper will outline the important parameters needed at an early stage to assess options and to help consider the balance between off-site and on-site treatment solutions, and the role of prior decontamination techniques. The paper also outlines the use of feasibility studies and demonstration projects that have been used to help customers prepare for decommissioning. The paper discusses physical, radiological and operational history data, Pro and Contra factors for on- and off-site treatment, the role of chemical decontamination prior to treatment, planning for off-site shipments as well as Studsvik experience This paper has an original focus upon the coming challenges of steam generator decommissioning and potential external treatment capacity constraints in the medium term. It also focuses on the potential during operations or initial shut-down to develop robust plans for steam generator management. (authors)

Analysis of steam generator loss-of-feedwater experiments with APROS and RELAP5/MOD3.1 computer codes

Energy Technology Data Exchange (ETDEWEB)

Virtanen, E.; Haapalehto, T. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Nuclear Energy, Lappeenranta (Finland)

1995-09-01

Three experiments were conducted to study the behavior of the new horizontal steam generator construction of the PACTEL test facility. In the experiments the secondary side coolant level was reduced stepwise. The experiments were calculated with two computer codes RELAP5/MOD3.1 and APROS version 2.11. A similar nodalization scheme was used for both codes to that the results may be compared. Only the steam generator was modelled and the rest of the facility was given as a boundary condition. The results show that both codes calculate well the behaviour of the primary side of the steam generator. On the secondary side both codes calculate lower steam temperatures in the upper part of the heat exchange tube bundle than was measured in the experiments.

The use of advanced scale conditioning agents for maintenance of the secondary side of nuclear plant steam generators

International Nuclear Information System (INIS)

Battaglia, P.J.; Rogosky, D.L.

2006-01-01

Maintaining the secondary side of steam generators within a pressurized water reactor (PWR) free of deposited corrosion products and corrosion-inducing contaminants is key to ensuring their long-term operation. New cleaning processes have become available to aid nuclear plant personnel in optimizing secondary side maintenance strategies. These strategies include both maintaining nuclear steam generators corrosion free while maintaining full power operation. The conference presentation will discuss ASCA use and the major field experience acquired in the last several years in the United States and in Japan. Hokkaido Electric, Dominion Engineering, Inc. and Westinghouse cosponsored the development of ASCAs for use in the Nuclear Utility industry, and all three are active in field use programs. Westinghouse owns the worldwide rights for ASCA implementation except in Japan where MHI and NEL have been granted licenses to apply ASCAs. Dominion Engineering Inc., owns the ASCA patents and performs the laboratory qualification testing associated with the ASCA programs, and Hokkaido Electric are joint patent holders for ASCAs and have been implementing their use at the Tomari plants for cleaning and thermal hydraulic performance enhancements. The specific experience discussed in the presentation will include: 1. Full Bundle Maintenance ASCAs at Vogtle Units 2 and 2 and Wolf Creek (USA). 2. Top of the Tubesheet ASCAs with high pressure sludge lancing at Wolf Creek and UEC at Vogtle Units 1 and 2 (USA). 3. Thermal Hydraulic Recovery and Maintenance ASCAs at the Hokkaido Electric Tomari Units 1 and 2 (Japan). (author)

Development of the visual inspection system for the top of the tube sheet in steam generators

International Nuclear Information System (INIS)

Kim, Gyung Sub; Choi, Sang Hoon; Kim, Ki Chul

2008-01-01

Steam Generators at Nuclear Power plants have a important function to isolate Radioactivity between the primary side radioactive fluid running through tubes and the secondary side with non-radioactive fluid through out of a tube bundle, in addition to a function of steam generation. Therefore, To obtain integrity of Steam Generator is really important for safety in the nuclear power plant. At the same time, sludge and foreign objects in steam generators are known as major sources causing the damage of SG tubes. But there is no way to prevent those coming to steam generators until now. Therefore, a periodic inspection and removal of those in steam generators is the only way for those Generally, Most of the Nuclear Power Plants have been inspecting visually every outage for the top of the tube sheet in which sludge and foreign objects lead to the buildup to know how these are

Mathematical modeling of a steam generator for sensor fault detection

International Nuclear Information System (INIS)

Prock, J.

1988-01-01

A dynamic model for a nuclear power plant steam generator (vertical, preheated, U-tube recirculation-type) is formulated as a sixth-order nonlinear system. The model integrates nodal mass and energy balances for the primary water, the U-tube metal and the secondary water and steam. The downcomer flow is determined by a static balance of momentum. The mathematical system is solved using transient input data from the Philippsburg 2 (FRG) nuclear power plant. The results of the calculation are compared with actual measured values. The proposed model provides a low-cost tool for the automatic control and simulation of the steam generating process. The ''parity-space'' algorithm is used to demonstrate the applicability of the mathematical model for sensor fault detection and identification purposes. This technique provides a powerful means of generating temporal analytical redundancy between sensor signals. It demonstrates good detection rates of sensor errors using relatively few steps of scanning time and allows the reconfiguration of faulty signals. (author)

Steam generator tube performance

International Nuclear Information System (INIS)

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

1982-04-01

The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1980. Tube defects occurred at 38% of the 97 reactors surveyed. This is a marginal improvement over 1979 when defects occurred at 41% of the reactors. The number of failed tubes was also lower, 0.14% of the tubes in service in 1980 compared with 0.20% of those in service in 1979. Analysis of the causes of these failures indicates that stress corrosion cracking was the leading failure mechanism. Reactors that used all-volatile treatment of secondary water, with or without full-flow condensate demineralization since start-up showed the lowest incidence of corrosion-related defects

Liquid metal steam generator

International Nuclear Information System (INIS)

Wolowodiuk, W.

1975-01-01

A liquid metal heated steam generator is described which in the event of a tube failure quickly exhausts out of the steam generator the products of the reaction between the water and the liquid metal. The steam is generated in a plurality of bayonet tubes which are heated by liquid metal flowing over them between an inner cylinder and an outer cylinder. The inner cylinder extends above the level of liquid metal but below the main tube sheet. A central pipe extends down into the inner cylinder with a centrifugal separator between it and the inner cylinder at its lower end and an involute deflector plate above the separator so that the products of a reaction between the liquid metal and the water will be deflected downwardly by the deflector plate and through the separator so that the liquid metal will flow outwardly and away from the central pipe through which the steam and gaseous reaction products are exhausted. (U.S.)

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)

Circumferential cracking of steam generator tubes

International Nuclear Information System (INIS)

Karwoski, K.J.

1997-04-01

On April 28, 1995, the U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 95-03, open-quote Circumferential Cracking of Steam Generator Tubes.close-quote GL 95-03 was issued to obtain information needed to verify licensee compliance with existing regulatory requirements regarding the integrity of steam generator tubes in domestic pressurized-water reactors (PWRs). This report briefly describes the design and function of domestic steam generators and summarizes the staff's assessment of the responses to GL 95-03. The report concludes with several observations related to steam generator operating experience. This report is intended to be representative of significant operating experience pertaining to circumferential cracking of steam generator tubes from April 1995 through December 1996. Operating experience prior to April 1995 is discussed throughout the report, as necessary, for completeness

Steam generator sludge removal apparatus

International Nuclear Information System (INIS)

Schafer, B.W.; Werner, C.E.; Klahn, F.C.

1992-01-01

The present invention relates to equipment for cleaning steam generators and in particular to a high pressure fluid lance for cleaning sludge off the steam generator tubes away from an open tube lane. 6 figs

Modeling and field studies of fouling in once-through steam generators

International Nuclear Information System (INIS)

Thompson, R.; Gaudreau, T.

1995-01-01

Efforts of the past 10 years to minimize fouling of the Crystal River-3 once-through steam generators are reviewed. The major focus has been on improving at-temperature pH control in the secondary cycle. Various concentrations of different pH control agents were tested in the field for hundreds of days to determine their effect on steam generator fouling. High concentrations of morpholine (50--100 ppm) in the feedwater were found to apparently produce de-fouling of the steam generators without an associated decrease in feedwater iron concentration as compared to that at lower levels of morpholine. Computer modeling of the pH(t) within the OTSG for the various chemistries tested indicates that the pH can change significantly with elevation within the steam generator by varying the pH control agent or its concentration. It is postulated that these variations in pH may change the surface charge of the tubes, tube support plates, and/or corrosion product particles in solution, to favor either deposition or repulsion of the particles, and thereby producing conditions that either favor fouling or de-fouling of the OTSG. Crystal River-3 experience indicates that corrosion product deposition and release processes inside the steam generator can be chemically manipulated to favor release, and thereby maximize plant performance, and delay or avoid costly hydraulic or chemical cleanings

The Evaluation of Steam Generator Level Measurement Model for OPR1000 Using RETRAN-3D

International Nuclear Information System (INIS)

Doo Yong Lee; Soon Joon Hong; Byung Chul Lee; Heok Soon Lim

2006-01-01

Steam generator level measurement is important factor for plant transient analyses using best estimate thermal hydraulic computer codes since the value of steam generator level is used for steam generator level control system and plant protection system. Because steam generator is in the saturation condition which includes steam and liquid together and is the place that heat exchange occurs from primary side to secondary side, computer codes are hard to calculate steam generator level realistically without appropriate level measurement model. In this paper, we prepare the steam generator models using RETRAN-3D that include geometry models, full range feedwater control system and five types of steam generator level measurement model. Five types of steam generator level measurement model consist of level measurement model using elevation difference in downcomer, 1D level measurement model using fluid mass, 1D level measurement model using fluid volume, 2D level measurement model using power and fluid mass, and 2D level measurement model using power and fluid volume. And we perform the evaluation of the capability of each steam generator level measurement model by simulating the real plant transient condition, the title is 'Reactor Trip by The Failure of The Deaerator Level Control Card of Ulchin Unit 3'. The comparison results between real plant data and RETRAN-3D analyses for each steam generator level measurement model show that 2D level measurement model using power and fluid mass or fluid volume has more realistic prediction capability compared with other level measurement models. (authors)

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

Analysis of steam generator loss-of-feedwater experiments with APROS and RELAP5/MOD3.1 computer codes

International Nuclear Information System (INIS)

Virtanen, E.; Haapalehto, T.; Kouhia, J.

1997-01-01

Three experiments were conducted to study the behaviour of the new horizontal steam generator construction of the PACTEL test facility. In the experiments the secondary side coolant level was reduced stepwise. The experiments were calculated with two computer codes RELAP5/MOD3.1 and APROS version 2.11. A similar nodalization scheme was used for both codes so that the results may be compared. Only the steam generator was modeled and the rest of the facility was given as a boundary condition. The results show that both codes calculate well the behaviour of the primary side of the steam generator. On the secondary side both codes calculate lower steam temperatures in the upper part of the heat exchange tube bundle than was measured in the experiments. (orig.)

Steam generator tube rupture (SGTR) scenarios

International Nuclear Information System (INIS)

Auvinen, A.; Jokiniemi, J.K.; Laehde, A.; Routamo, T.; Lundstroem, P.; Tuomisto, H.; Dienstbier, J.; Guentay, S.; Suckow, D.; Dehbi, A.; Slootman, M.; Herranz, L.; Peyres, V.; Polo, J.

2005-01-01

The steam generator tube rupture (SGTR) scenarios project was carried out in the EU 5th framework programme in the field of nuclear safety during years 2000-2002. The first objective of the project was to generate a comprehensive database on fission product retention in a steam generator. The second objective was to verify and develop predictive models to support accident management interventions in steam generator tube rupture sequences, which either directly lead to severe accident conditions or are induced by other sequences leading to severe accidents. The models developed for fission product retention were to be included in severe accident codes. In addition, it was shown that existing models for turbulent deposition, which is the dominating deposition mechanism in dry conditions and at high flow rates, contain large uncertainties. The results of the project are applicable to various pressurised water reactors, including vertical steam generators (western PWR) and horizontal steam generators (VVER)

Minimize corrosion degradation of steam generator tube materials

International Nuclear Information System (INIS)

Lu, Y.

2006-01-01

As part of a coordinated program, AECL is developing a set of tools to aid with the prediction and management of steam generator performance. Although stress corrosion cracking (of Alloy 800) has not been detected in any operating steam generator, for life management it is necessary to develop mechanistic models to predict the conditions under which stress corrosion cracking is plausible. Experimental data suggest that all steam generator tube materials are susceptible to corrosion degradation under some specific off-specification conditions. The tolerance to the chemistry upset for each steam generator tube alloy is different. Electrochemical corrosion behaviors of major steam generator tube alloys were studied under the plausible aggressive crevice chemistry conditions. The potential hazardous conditions leading to steam generator tube degradation and the conditions, which can minimize steam generator tube degradation have been determined. Recommended electrochemical corrosion potential/pH zones were defined for all major steam generator tube materials, including Alloys 600, 800, 690 and 400, under CANDU steam generator operating and startup conditions. Stress corrosion cracking tests and accelerated corrosion tests were carried out to verify and revise the recommended electrochemical corrosion potential/pH zones. Based on this information, utilities can prevent steam generator material degradation surprises by appropriate steam generator water chemistry management and increase the reliability of nuclear power generating stations. (author)

Internal oxidation as a mechanism for steam generator tube degradation

Energy Technology Data Exchange (ETDEWEB)

Gendron, T.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Scott, P.M. [Framatome, Paris (France); Bruemmer, S.M. [Pacific Northwest National Laboratory, Richland, WA (United States); Thomas, L.E. [Washington State Univ., School of Mechanical and Materials Engineering, Pullman, WA (United States)

1999-12-01

Internal oxidation has been proposed as a plausible mechanism for intergranular stress-corrosion cracking (IGSCC) of alloy 600 steam generator tubing. This theory can reconcile the main thermodynamic and kinetic characteristics of the observed cracking in hydrogenated primary water. Although secondary-side IG attack or IGSCC is commonly attributed to the presence of strong, caustic or acidic solutions, more recent evidence suggests that this degradation takes place in a near neutral environment, possibly dry polluted steam. As a result, internal oxidation is also a feasible mechanism for secondary side degradation. The present paper reviews experimental work conducted in an attempt to determine the validity of this mechanism. The consequences for the expected behaviour of alloys 690 and 800 replacement materials are also described. (author)

Internal oxidation as a mechanism for steam generator tube degradation

Energy Technology Data Exchange (ETDEWEB)

Gendron, T.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Scott, P.M. [Framatome, Paris (France); Bruemmer, S.M. [Pacific Northwest National Lab., Richland, Washington (United States); Thomas, L.E. [Washington State Univ., School of Mechanical and Materials Engineering, Pullman, WA (United States)

1998-07-01

Internal oxidation has been proposed as a plausible mechanism for intergranular stress corrosion cracking (IGSCC) of alloy 600 steam generator tubing. This theory can reconcile the main thermodynamic and kinetic characteristics of the observed cracking in hydrogenated primary water. Although secondary side IG attack or IGSCC is commonly attributed to the presence of strong caustic or acidic solutions, more recent evidence suggests that this degradation takes place in a near-neutral environment, possibly dry polluted steam. As a result, internal oxidation is also a feasible mechanism for secondary side degradation. The present paper reviews experimental work carried out in an attempt to determine the validity of this mechanism. The consequences for the expected behaviour of alloys 690 and 800 replacement materials are also described. (author)

Internal oxidation as a mechanism for steam generator tube degradation

International Nuclear Information System (INIS)

Gendron, T.S.; Scott, P.M.; Bruemmer, S.M.; Thomas, L.E.

1998-01-01

Internal oxidation has been proposed as a plausible mechanism for intergranular stress corrosion cracking (IGSCC) of alloy 600 steam generator tubing. This theory can reconcile the main thermodynamic and kinetic characteristics of the observed cracking in hydrogenated primary water. Although secondary side IG attack or IGSCC is commonly attributed to the presence of strong caustic or acidic solutions, more recent evidence suggests that this degradation takes place in a near-neutral environment, possibly dry polluted steam. As a result, internal oxidation is also a feasible mechanism for secondary side degradation. The present paper reviews experimental work carried out in an attempt to determine the validity of this mechanism. The consequences for the expected behaviour of alloys 690 and 800 replacement materials are also described. (author)

Hard sludge formation in modern steam generators of nuclear power plants. Formation, risks and mitigation

International Nuclear Information System (INIS)

Strohmer, F.

2014-01-01

In recent years modern steam generators with triangular pitch tube bundle geometry have experienced damage caused by hard sludge formation on top of the tube sheet and denting. The effect can lead to a limitation of the modern steam generators’ lifetime. The current publication shows reasons for the generation of hard sludge formation. Moreover, it describes the risk arising from hard sludge formation for the concerned steam generators and the mitigation of the problem. The main factors contributing to the formation of hard sludge are: the amount of corrosion product ingress into the steam generators, hard sludge formation favouring impurities and, skipped maintenance applications during outages. The main damaging mechanism of denting that can arise under certain secondary side conditions from hard sludge is explained. For steam generator tube denting, aggressive, oxidizing conditions have to be established in crevices beneath the hard sludge piles. Severely dented tubes are sensitive toward outer diameter stress corrosion cracking (ODSCC). The denting and ODSCC mechanism is explained. In addition, a proactive long-term maintenance strategy to avoid the formation of hard sludge piles will be shown. The strategy is based on a reduction of the corrosion product ingress into the steam generator's secondary side, and on the regular removal of deposits from the tube sheet and from the entire upper bundle area by latest mechanical cleaning methods. For hard deposits - formed either by silicates or long term hardened corrosion products, which, in the past, could not be removed by chemical or mechanical means - a new, simple, mechanical cleaning method is presented. This method can be used during the normal time frame of an outage and allows the restart of the unit with clean steam generator tube sheets. This mitigates the tendency to form hard sludge and denting in the long term. (author)

Regulation of ageing steam generators

International Nuclear Information System (INIS)

Jarman, B.L.; Grant, I.M.; Garg, R.

1998-01-01

Recent years have seen leaks and shutdowns of Canadian CANDU plants due to steam generator tube degradation by mechanisms including stress corrosion cracking, fretting and pitting. Failure of a single steam generator tube, or even a few tubes, would not be a serious safety related event in a CANDU reactor. The leakage from a ruptured tube is within the makeup capacity of the primary heat transport system, so that as long as the operator takes the correct actions, the off-site consequences will be negligible. However, assurance that no tubes deteriorate to the point where their integrity could be seriously breached as result of potential accidents, and that any leakage caused by such an accident will be small enough to be inconsequential, can only be obtained through detailed monitoring and management of steam generator condition. This paper presents the AECB's current approach and future regulatory directions regarding ageing steam generators. (author)

Tube sheet design for PFBR steam generator

International Nuclear Information System (INIS)

Chellapandi, P.; Chetal, S.C.; Bhoje, S.B.

1991-01-01

Top and bottom tube sheets of PFBR Steam Generators have been analysed with 3D and axisymmetric models using CASTEM Programs. Analysis indicates that the effects of piping reactions at the inlet/outlet nozzles on the primary stresses in the tube sheets are negligible and the asymmetricity of the deformation pattern introduced in the tube sheet by the presence of inlet/outlet and manhole nozzles is insignificant. The minimum tube sheet thicknesses for evaporator and reheater are 135 mm and 75 mm respectively. Further analysis has indicated the minimum fillet radius at the junction of tube sheet and dished end should be 20 mm. Simplified methodology has been developed to arrive at the number of thermal baffles required to protect the tube sheet against fatigue damage due to thermal transient. This method has been applied to PFBR steam generators to determine the required number of thermal baffles. For protecting the bottom tube sheet of evaporator against the thermal shock due to feed water and secondary pump trip, one thermal shield is found to be sufficient. Further analysis is required to decide upon the actual number to take care of the severe thermal transient, following the event of sudden dumping of water/steam, immediately after the sodium-water reaction. (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.

Application of perturbation methods for sensitivity analysis for nuclear power plant steam generators

International Nuclear Information System (INIS)

Gurjao, Emir Candeia

1996-02-01

The differential and GPT (Generalized Perturbation Theory) formalisms of the Perturbation Theory were applied in this work to a simplified U-tubes steam generator model to perform sensitivity analysis. The adjoint and importance equations, with the corresponding expressions for the sensitivity coefficients, were derived for this steam generator model. The system was numerically was numerically solved in a Fortran program, called GEVADJ, in order to calculate the sensitivity coefficients. A transient loss of forced primary coolant in the nuclear power plant Angra-1 was used as example case. The average and final values of functionals: secondary pressure and enthalpy were studied in relation to changes in the secondary feedwater flow, enthalpy and total volume in secondary circuit. Absolute variations in the above functionals were calculated using the perturbative methods, considering the variations in the feedwater flow and total secondary volume. Comparison with the same variations obtained via direct model showed in general good agreement, demonstrating the potentiality of perturbative methods for sensitivity analysis of nuclear systems. (author)

Numerical analysis of APR1400 Steam Generator by CUPID/MARS heat structure coupling

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Ryong Lee; Lee, Seung Jun; Pakr, Ik Kyu; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of); Cho, Hyoung Kyu [Seoul National University, Seoul (Korea, Republic of)

2015-05-15

To design and analyze steam generators, many computer codes have been developed and used around the world. In this study, the coupled CUPID and MARS code was used for the simulation of boiler side of the PWR steam generator. This paper presents the description of the coupling method, validation for porous media approach against the rod bundle experiment and the preliminary simulation results of PWR steam generator using the coupled code. In the present study, the multi-scale thermal-hydraulic analysis method using the coupled CUPID/MARS code was applied for the simulation of the steam generator. The primary side of the steam generator and other RCS was simulated by MARS and the secondary side was calculated by CUPID with porous media approach. For coupled simulation, the porous medium was applied in order to take into account the effect of the U-tube bundle and other supporting structure which play a role to be a flow resistance. More realistic physical model such as moisture separator slug behavior should be developed for the near future. The application of the coupled simulation should be extended to the accident scenario.

Steam generator and circulator model for the HELAP code

International Nuclear Information System (INIS)

Ludewig, H.

1975-07-01

An outline is presented of the work carried out in the 1974 fiscal year on the GCFBR safety research project consisting of the development of improved steam generator and circulator (steam turbine driven helium compressor) models which will eventually be inserted in the HELAP (1) code. Furthermore, a code was developed which will be used to generate steady state input for the primary and secondary sides of the steam generator. The following conclusions and suggestions for further work are made: (1) The steam-generator and circulator model are consistent with the volume and junction layout used in HELAP, (2) with minor changes these models, when incorporated in HELAP, could be used to simulate a direct cycle plant, (3) an explicit control valve model is still to be developed and would be very desirable to control the flow to the turbine during a transient (initially this flow will be controlled by using the existing check valve model); (4) the friction factor in the laminar flow region is computed inaccurately, this might cause significant errors in loss-of-flow accidents; and (5) it is felt that HELAP will still use a large amount of computer time and will thus be limited to design basis accidents without scram or loss of flow transients with and without scram. Finally it may also be used as a test bed for the development of prototype component models which would be incorporated in a more sophisticated system code, developed specifically for GCFBR's

Numerical analysis of gas-liquid two-phase flow in secondary side of steam generator

Energy Technology Data Exchange (ETDEWEB)

Murase, Michio; Nakamura, Akira; Yagi, Yoshinori [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2002-09-01

The steam generator (SG) in a pressurized water reactor (PWR) is an important two-phase flow component as the boundary between the primary loop and the secondary loop. In this study, we performed gas-liquid two-phase flow analyses for SG reliability tests conduced by Nuclear Power Engineering Corporation (NUPEC) using the two-fluid model of a thermal-hydraulic computer code PHOENICS. In order to calculate the location of the boiling initiation accurately, detailed inputs were required for the friction coefficients affecting the velocity distribution and the heat transfer distribution. However, the velocity and heat transfer distributions did not greatly affect the void fractions in the upper region of the heat transfer tubes. The calculated void fractions agreed with the measured values within 4% as the local average and within 2% as an average in a cross-section, except the region of low void fractions. (author)

Simulating Porous Magnetite Layer Deposited on Alloy 690TT Steam Generator Tubes.

Science.gov (United States)

Jeon, Soon-Hyeok; Son, Yeong-Ho; Choi, Won-Ik; Song, Geun Dong; Hur, Do Haeng

2018-01-02

In nuclear power plants, the main corrosion product that is deposited on the outside of steam generator tubes is porous magnetite. The objective of this study was to simulate porous magnetite that is deposited on thermally treated (TT) Alloy 690 steam generator tubes. A magnetite layer was electrodeposited on an Alloy 690TT substrate in an Fe(III)-triethanolamine solution. After electrodeposition, the dense magnetite layer was immersed to simulate porous magnetite deposits in alkaline solution for 50 days at room temperature. The dense morphology of the magnetite layer was changed to a porous structure by reductive dissolution reaction. The simulated porous magnetite layer was compared with flakes of steam generator tubes, which were collected from the secondary water system of a real nuclear power plant during sludge lancing. Possible nuclear research applications using simulated porous magnetite specimens are also proposed.

The impact of NPP Krsko steam generator tube plugging on minimum DNBR at nominal conditions

International Nuclear Information System (INIS)

Lajtman, S.

1996-01-01

Typically, steam generator tube plugging (SGTP) both decreases the reactor coolant system (RCS) flow rate and the heat transfer surface area of the steam generator. At a constant thermal power and vessel outlet temperature, as tube plugging increases, the vessel average temperature, vessel inlet temperature and steam generator secondary side steam pressure decrease. This paper presents the analysis of impact of SGTP on Minimum Departure from Nucleate Boiling Ratio (MDNBR) at NPP Krsko (NEK), using the Improved Thermal Design Procedure (ITDP), WRB-1 correlation, and COBRA-III-C computer code. No credit was given to high plugging percentage region power reduction resulting from turbine volumetric flow limitations. MDNBR is found to be decreasing with increasing plugging, but not under the limiting values. (author)

Ultrasonic downcomer flow measurements for recirculating steam generators

Energy Technology Data Exchange (ETDEWEB)

Janzen, Victor, E-mail: Victor.Janzen@cnl.ca [Canadian Nuclear Laboratories, Chalk River, ON, Canada K0 J 1J0 (Canada); Luloff, Brian [Canadian Nuclear Laboratories, Chalk River, ON, Canada K0 J 1J0 (Canada); Sedman, Ken [Nuclear Safety Analysis & Support Department, Bruce Power, Toronto, ON, Canada M5G 1X6 (Canada)

2015-08-15

Highlights: • Measuring recirculating flow in nuclear steam generators provides useful information. • Flow measurements shed light on component performance and degradation mechanisms. • Commonly used ultrasonic technology and application methods are described. • Results of measurements at several power reactors are summarized. • Potential improvements in reliability and flexibility of application are suggested. - Abstract: Measurements of downcomer flow in nuclear steam generators can provide unique fitness for service and performance indicators related to overall thermalhydraulic performance, safety related secondary-side setpoints and certain forms of degradation. This paper reviews the benefits of downcomer-flow measurements to nuclear power–plant operators, and describes methods that are commonly used. It summarizes the history and state-of-the-art of the most widely used technology, non-intrusive ultrasonic systems, including field applications at several nuclear power plants. It also describes the technical challenges that remain, and summarizes recent technical developments and future improvements.

Small leak detection requirements for SNR-300 steam generator operation

International Nuclear Information System (INIS)

Dumm, K.

1975-01-01

The leak detection philosophy of the SNR-300 steam generators is described. Due to operational demands small leaks have already to be detectable at a stage where secondary tube damage has not yet occurred. The requirements on a leak detection instrument are developed and the instrument itself is described. (author)

Small leak detection requirements for SNR-300 steam generator operation

Energy Technology Data Exchange (ETDEWEB)

Dumm, K

1975-07-01

The leak detection philosophy of the SNR-300 steam generators is described. Due to operational demands small leaks have already to be detectable at a stage where secondary tube damage has not yet occurred. The requirements on a leak detection instrument are developed and the instrument itself is described. (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)

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

Failures of fine tubes of steam generators and the essential defects

International Nuclear Information System (INIS)

Kawano, Shinji; Ebisawa, Toru; Sato, Susumu.

1976-01-01

Light water reactors were sold to Japan as their economy and safety have been established, but the average availability of 11 reactors in Japan during 7 year operation is only 53%, and it is being proved that there are questions in the safety and economy. In this report, the failures of fine tubes of steam generators are discussed from the standpoint of the corrosion of materials. First, the functions and construction of the fine tubes of steam generators in PWRs are explained. The failures of the fine tubes of steam generators became frequent since the beginning of 1970s as large capacity nuclear power stations have started the operation. When the fine tubes are pierced with holes during operation and the radioactivity in primary coolant leaks into secondary coolant, it is detected with radioactivity monitors. In order to find out the broken tubes, eddy current flaw detectors are used, and the tubes on which flaws were detected we plugged by explosion welding. In these works, many manual operations are included, and the radiation exposure of workers and the difficulties in the operations are the problems. The cases of the tube failures in Japan and foreign countries, the causes and the countermeasures are described. Chemical corrosion, thermal stress cycle, shaving off due to eddy flow, and stress corrosion are the probable causes. The safety of steam generators is essentially in extremely poor state. The seriousness of the tube failures in steam generators is emphasized. (Kako, I.)

Equipment for secondary water distribution in heat exchanger, especially saturated steam generator for nuclear power plants and heat plants

Energy Technology Data Exchange (ETDEWEB)

Riman, J; Manek, O; Rybnicek, J

1979-09-15

A special structure consisting of a system of channels and a distribution plate with ports in-built above the tube-plate of a vertical-type steam generator prevents secondary water vaporization in the space above the tube-plate and thus also salt and sludge sedimentation which causes increased corrosion of heat transfer tubes. The size of the distribution plate ports is variable in the radial direction. The distribution plate is divided by means of the system of channels into at least two parts. The middle section of each part is of the through-flow type.

Moisture separator for steam generator level measurement system

International Nuclear Information System (INIS)

Cantineau, B.J.

1987-01-01

A steam generator level measurement system having a reference leg which is kept full of water by a condensation pot, has a liquid/steam separator in the connecting line between the condensation pot and the steam phase in the steam generator to remove excess liquid from the steam externally of the steam generator. This ensures that the connecting line does not become blocked. The separator pot has an expansion chamber which slows down the velocity of the steam/liquid mixture to aid in separation, and a baffle, to avoid liquid flow into the line connected to the condensate pot. Liquid separated is returned to the steam generator below the water level through a drain line. (author)

Passive system with steam-water injector for emergency supply of NPP steam generators

International Nuclear Information System (INIS)

Il'chenko, A.G.; Strakhov, A.N.; Magnitskij, D.N.

2009-01-01

The calculation results of reliability indicators of emergency power supply system and emergency feed-water supply system of serial WWER-1000 unit are presented. To ensure safe water supply to steam generators during station blackout it was suggested using additional passive emergency feed-water system with a steam-water injector working on steam generators dump steam. Calculated analysis of steam-water injector operating capacity was conducted at variable parameters of steam at the entrance to injector, corresponding to various moments of time from the beginning of steam-and-water damping [ru

Thermal-hydraulics in recirculating steam generators

International Nuclear Information System (INIS)

Carver, M.B.; Carlucci, L.N.; Inch, W.W.R.

1981-04-01

This manual describes the THIRST code and its use in computing three-dimensional two-phase flow and heat transfer in a steam generator under steady state operation. The manual is intended primarily to facilitate the application of the code to the analysis of steam generators typical of CANDU nuclear stations. Application to other steam generator designs is also discussed. Details of the assumptions used to formulate the model and to implement the numerical solution are also included

Tests and analysis on steam generator tube failure propagation

International Nuclear Information System (INIS)

Tanabe, Hiromi

1990-01-01

The understanding of leak enlargement and failure propagation behavior is essential to select a design basis leak (DBL) of LMFBR steam generators. Therefore, various series of experiments, such as self-enlargement tests, target wastage tests, failure propagation tests were conducted in a wide range of leak using test facilities of SWAT at PNC/OEC. Especially, in the large leak tests, potential of overheating failure was investigated under a prototypical steam cooling condition inside target tubes. In the small leak, the difference of wastage resistivity was clarified among several tube materials such as 9-chrome steels. In regard to an analytical approach, a computer code LEAP (Leak Enlargement and Propagation) was developed on the basis of all of these experimental results. The code was used to validate the previously selected DBL of the prototype reactor, Monju, steam generator. This approach proved to be successful in spite of somewhat over-conservatism in the analysis. Moreover, LEAP clarified the effectiveness of a rapid steam dump and an enhanced leak detection system. The code improvement toward a realistic analysis is desired, however, to lessen the DBL for a future large plant and then the re-evaluation of the experimental data such as the size of secondary failure is under way. (author). 4 refs, 8 figs, 1 tab

Steam generator life cycle management: Ontario Power Generation (OPG) experience

International Nuclear Information System (INIS)

Maruska, C.C.

2002-01-01

A systematic managed process for steam generators has been implemented at Ontario Power Generation (OPG) nuclear stations for the past several years. One of the key requirements of this managed process is to have in place long range Steam Generator Life Cycle Management (SG LCM) plans for each unit. The primary goal of these plans is to maximize the value of the nuclear facility through safe and reliable steam generator operation over the expected life of the units. The SG LCM plans integrate and schedule all steam generator actions such as inspection, operation, maintenance, modifications, repairs, assessments, R and D, performance monitoring and feedback. This paper discusses OPG steam generator life cycle management experience to date, including successes, failures and how lessons learned have been re-applied. The discussion includes relevant examples from each of the operating stations: Pickering B and Darlington. It also includes some of the experience and lessons learned from the activities carried out to refurbish the steam generators at Pickering A after several years in long term lay-up. The paper is structured along the various degradation modes that have been observed to date at these sites, including monitoring and mitigating actions taken and future plans. (author)

Non-polluting steam generators with fluidized-bed furnaces

Energy Technology Data Exchange (ETDEWEB)

Brandes, H [Deutsche Babcock A.G., Oberhausen (Germany, F.R.)

1979-07-01

The author reports on a 35 MW steam generator with hard coal fluidized-bed furnace a planned 35 MW steam generator with flotation-dirt fluidized-bed furnace, and on planned steam generators for fluidized-bed firing of hard coal up to a steam power of about 200 MW.

CAREM-25 Steam Generator Stability Analysis

International Nuclear Information System (INIS)

Rabiti, A.; Delmastro, D.

2003-01-01

In this work the stability of a once-through CAREM-25 steam generator is analyzed.A fix nodes numerical model, that allows the modelling of the liquid, two-phase and superheated steam zones, is implemented.This model was checked against a mobile finite elements model under saturated steam conditions at the channel exit and a good agreement was obtained.Finally the stability of a CAREM steam generator is studied and the range of in let restrictions that a assure the system stability is analyzed

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.)

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

Darlington steam generator life assurance program

International Nuclear Information System (INIS)

Jelinski, E.; Dymarski, M.; Maruska, C.; Cartar, E.

1995-01-01

The Darlington Nuclear Generating Station belonging to Ontario Hydro is one of the most modern and advanced nuclear generating stations in the world. Four reactor units each generate 881 net MW, enough to provide power to a major city, and representing approximately 20% of the Ontario grid. The nuclear generating capacity in Ontario represents approximately 60% of the grid. In order to look after this major asset, many proactive preventative and predictive maintenance programs are being put in place. The steam generators are a major component in any power plant. World wide experience shows that nuclear steam generators require specialized attention to ensure reliable operation over the station life. This paper describes the Darlington steam generator life assurance program in terms of degradation identification, monitoring and management. The requirements for chemistry control, surveillance of process parameters, surveillance of inspection parameters, and the integration of preventative and predictive maintenance programs such as water lancing, chemical cleaning, RIHT monitoring, and other diagnostics to enhance our understanding of life management issues are identified and discussed. We conclude that we have advanced proactive activities to avoid and to minimize many of the problems affecting other steam generators. An effective steam generator maintenance program must expand the knowledge horizon to understand life limiting processes and to analyze and synthesize observations with theory. (author)

A two-fluid two-phase model for thermal-hydraulic analysis of a U-tube steam generator

International Nuclear Information System (INIS)

Hung, Huanjen; Chieng, Chingchang; Pei, Baushei; Wang, Songfeng

1993-01-01

The Advanced Thermal-Hydraulic Analysis Code for Nuclear Steam Generators (ATHANS) was developed on the basis of the THERMIT-UTSG computer code for U-tube steam generators. The main features of the ATHANS model are as follows: (a) the equations are solved in cylindrical coordinates, (b) the number and the arrangement of the control volumes inside the steam generator can be chosen by the user, (c) the virtual mass effect is incorporated, and (d) the conjugate gradient squared method is employed to accelerate and improve the numerical convergence. The performance of the model is successfully validated by comparison with the test data from a Westinghouse model F steam generator at the Maanshan nuclear power plant. Better agreement with the test data can be obtained by a finer grid system using a cylindrical coordinate system and the virtual mass effect. With these advanced features, ATHANS provides the basic framework for further studies on the problems of steam generators, such as analyses of secondary-side corrosion and tube ruptures

Materials choices for the advanced LWR steam generators

International Nuclear Information System (INIS)

Paine, J.P.N.; Shoemaker, C.E.; McIlree, A.R.

1987-01-01

Current light water reactor (LWR) steam generators have been affected by a variety of corrosion and mechanical damage degradation mechanisms. Included are wear caused by tube vibration, intergranular corrosion, pitting, and thinning or wastage of the steam generator tubing and accelerated corrosion of carbon steel supports (denting). The Electric Power Research Institute (EPRI) and the Steam Generator Owners Groups (I, II) have sponsored laboratory and field studies to provide ameliorative actions for the majority of the damage forms experienced to date. Some of the current corrosion mechanisms are aggravated or caused by unique materials choices or materials interactions. New materials have been proposed and at least partially qualified for use in replacement model steam generators, including an advanced LWR design. In so far as possible, the materials choices for the advanced LWR steam generator avoid the corrosion pitfalls seemingly inherent in the current designs. The EPRI Steam Generator Project staff has recommended materials and design choices for a new steam generator. Based on these recommendations we believe that the advanced LWR steam generators will be much less affected by corrosion and mechanical damage mechanisms than are now experienced

PWR steam generator chemical cleaning, Phase I. Final report

International Nuclear Information System (INIS)

Rothstein, S.

1978-07-01

United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the search sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI

Wolsong 3 and 4 steam generator tube inspection

International Nuclear Information System (INIS)

Jang, Kyoung Sik; Son, Tai Bong; Kwon, Dong Ki; Choi, Jin Hyuk

2001-01-01

During the pre-service inspection for Wolsong unit 3 and 4 in 1997/1998 respectively, 17 distorted roll transition indications (over expanded beyond tubesheet secondary face) were identified at the unit 4 (S/G B, D). Six(6) tubes out of these tubes were plugged in 1998. However the first periodic inspection identified additional 110 indications in 1999 and 2000. The additionally identified 110 indication call, not reported at the pre-service inspection, are; 2 not-finally-expanded-tubes and 108 distorted roll transition tubes. Design limit of each steam generator tube plugging is 6.4.%. Plugging was performed by the steam generator manufacturer under the warranty. When distorted roll transition indications were first identified on the unit 4 in 1998 the degree of over-expansion was measured using an inner dial-gage to make the disposition of nonconformance report. 2 Not-finally-expanded-tubes were plugged and 10 tubes out of 108 distorted roll transition tubes were also plugged as a preventive measure

Strategic maintenance plan for Cernavoda steam generators

International Nuclear Information System (INIS)

Cicerone, T.; Dhar, D.; VandenBerg, J.P.

2002-01-01

Steam generators are among the most important pieces of equipment in a nuclear power plant. They are required full time during the plant operation and obviously no redundancy exists. Past experience has shown that those utilities which implemented comprehensive steam generator inspection and maintenance programs and strict water chemistry controls, have had good steam generator performance that supports good overall plant performance. The purpose of this paper is to discuss a strategic Life Management and Operational-monitoring program for the Cernavoda steam generators. The program is first of all to develop a base of expertise for the management of the steam generator condition; and that is to be supported by a program of actions to be accomplished over time to assess their condition, to take measures to avoid degradation and to provide for inspections, cleaning and modifications as necessary. (author)

Vertical steam generator

International Nuclear Information System (INIS)

Cuda, F.; Kondr, M.; Kresta, M.; Kusak, V.; Manek, O.; Turon, S.

1982-01-01

A vertical steam generator for nuclear power plants and dual purpose power plants consists of a cylindrical vessel in which are placed heating tubes in the form upside-down U. The heating tubes lead to the jacket of the cylindrical collector placed in the lower part of the steam generator perpendicularly to its vertical axis. The cylindrical collector is divided by a longitudinal partition into the inlet and outlet primary water sections of the heating tubes. One ends of the heating tube leads to the jacket of the collector for primary water feeding and the second ends of the heating tubes into the jacket of the collector which feeds and offtakes primary water from the heating tubes. (B.S.)

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

Development of KALANS to remove sludge of steam generator

International Nuclear Information System (INIS)

Jeong, W. T.; Yang, J. S.; You, G. W.; Sung, G. B.; Kim, B. Y.

1999-01-01

CECIL -4 lancing system by Foster-Miller has been used for the secondary side landing of Westinghouse type steam generators (YGN No. 1,2 and Kori No. 1,2,3,4) in Korean nuclear power plants. Shared use of only one lancing system for the 6 nuclear power plants required frequent transportation of the system from Nori to Youngkwang which may cause malfunction of sophisticated control equipment. Sometimes, the overlap in preventive maintenance schedule of nuclear power plants prevented the plant maintenance personnel from doing sufficient lancing work. Purchasing a new lancing system may be a candidate for this problem, but the price tag is surprisingly high. Furthermore, prompt maintenance was another serious problem because the manufacturer is far away in a foreign country. Last year, the steam generator of Kori No. 1 was changed to Δ60. The difference in the tube arrangement of the new steam generator made it impossible to use the CECIL system. Sufficient technological background of our development team and the conclusion that the development project is economically feasible made us to commence the design and fabrication of 'KEPCO Advanced LANcing System - KALANS.' This paper presents the state of the art on lancing system and the developed concept and design of KALANS

The casebook of technical presentation on a steam generator

International Nuclear Information System (INIS)

1986-05-01

This casebook consists of seven presentations, which are measures and experience of maintenance of water quality in PWR generator, corrosion in steam generator, safely evaluation by management and closing in steam generator, testing of eddy current in steam generator, unsettled problems of safety in steam generator and maintenance of water quality in PWR generator.

Analysis and qualification of steam generator relief valves (BRU-A)

International Nuclear Information System (INIS)

Lathuile, C.; Serre, J. L.

1997-01-01

This paper presents a general overview of improvements foreseen in the frame of Safety Measures S01 and S10 in order to prevent and mitigate consequences of a large primary to secondary leakage. Among these improvements, a more detailed description of methodology and results relative to Steam Generator Relief Valves (BRU-A) qualification tests is presented. (author)

Model of reverse steam generator

International Nuclear Information System (INIS)

Malasek, V.; Manek, O.; Masek, V.; Riman, J.

1987-01-01

The claim of Czechoslovak discovery no. 239272 is a model designed for the verification of the properties of a reverse steam generator during the penetration of water, steam-water mixture or steam into liquid metal flowing inside the heat exchange tubes. The design may primarily be used for steam generators with a built-in inter-tube structure. The model is provided with several injection devices configured in different heat exchange tubes, spaced at different distances along the model axis. The design consists in that between the pressure and the circumferential casings there are transverse partitions and that in one chamber consisting of the circumferential casings, pressure casing and two adjoining partitions there is only one passage of the injection device through the inter-tube space. (Z.M.). 1 fig

Predicting steam generator crevice chemistry

International Nuclear Information System (INIS)

Burton, G.; Strati, G.

2006-01-01

'Full text:' Corrosion of steam cycle components produces insoluble material, mostly iron oxides, that are transported to the steam generator (SG) via the feedwater and deposited on internal surfaces such as the tubes, tube support plates and the tubesheet. The build up of these corrosion products over time can lead to regions of restricted flow with water chemistry that may be significantly different, and potentially more corrosive to SG tube material, than the bulk steam generator water chemistry. The aim of the present work is to predict SG crevice chemistry using experimentation and modelling as part of AECL's overall strategy for steam generator life management. Hideout-return experiments are performed under CANDU steam generator conditions to assess the accumulation of impurities in hideout, and return from, model crevices. The results are used to validate the ChemSolv model that predicts steam generator crevice impurity concentrations, and high temperature pH, based on process parameters (e.g., heat flux, primary side temperature) and blowdown water chemistry. The model has been incorporated into ChemAND, AECL's system health monitoring software for chemistry monitoring, analysis and diagnostics that has been installed at two domestic and one international CANDU station. ChemAND provides the station chemists with the only method to predict SG crevice chemistry. In one recent application, the software has been used to evaluate the crevice chemistry based on the elevated, but balanced, SG bulk water impurity concentrations present during reactor startup, in order to reduce hold times. The present paper will describe recent hideout-return experiments that are used for the validation of the ChemSolv model, station experience using the software, and improvements to predict the crevice electrochemical potential that will permit station staff to ensure that the SG tubes are in the 'safe operating zone' predicted by Lu (AECL). (author)

Development of a steam generator lancing system

International Nuclear Information System (INIS)

Jeong, Woo-Tae; Kim, Seok-Tae; Hong, Sung-Yull

2006-01-01

It is recommended to clean steam generators of nuclear power plants during plant outages. Under normal operations, sludge is created and constantly accumulates in the steam generators. The constituents of this sludge are different depending on each power plant characteristics. The sludge of the Kori Unit 1 steam generator, for example, was found to be composed of 93% ferrous oxide, 3% carbon and 1% of silica oxide and nickel oxide each. The research to develop a lancing system that would remove sludge deposits from the tubesheet of a steam generator was started in 1998 by the Korea Electric Power Research Institute (KEPRI) of the Korea Electric Power Corporation (KEPCO). The first commercial domestic lancing system in Korea, and KALANS-I Lancing System, was completed in 2000 for Kori Unit 1 for cleaning the tubesheet of its Westinghouse Delta-60 steam generator. Thereafter, the success of the development and site implementation of the KALANS-I lancing system for YGN Units 1 and 2 and Ulchin Units 3 and 4 was also realized in 2004 for sludge removal at those sites. The upper bundle cleaning system for Westinghouse model F steam generators is now under development

Method to detect steam generator tube leakage

International Nuclear Information System (INIS)

Watabe, Kiyomi

1994-01-01

It is important for plant operation to detect minor leakages from the steam generator tube at an early stage, thus, leakage detection has been performed using a condenser air ejector gas monitor and a steam generator blow down monitor, etc. In this study highly-sensitive main steam line monitors have been developed in order to identify leakages in the steam generator more quickly and accurately. The performance of the monitors was verified and the demonstration test at the actual plant was conducted for their intended application to the plants. (author)

Future aspects for liquid metal heated steam generators

International Nuclear Information System (INIS)

Jansing, W.; Ratzel, W.; Vinzens, K.

1975-01-01

The present status of steam generators is shown. The experience gained until now is expressed in form of basic points. The most important design criteria for steam generator systems are outlined. On the basis of these design criteria, two possible steam generator concepts are shown. Costs in relationship to the repair concepts of two modular steam generators (thermal output 156 and 625 MW) and a pool design of 625 MW are compared. (author)

Future aspects for liquid metal heated steam generators

Energy Technology Data Exchange (ETDEWEB)

Jansing, W; Ratzel, W; Vinzens, K

1975-07-01

The present status of steam generators is shown. The experience gained until now is expressed in form of basic points. The most important design criteria for steam generator systems are outlined. On the basis of these design criteria, two possible steam generator concepts are shown. Costs in relationship to the repair concepts of two modular steam generators (thermal output 156 and 625 MW) and a pool design of 625 MW are compared. (author)

Comments on US LMFBR steam generator base technology

International Nuclear Information System (INIS)

Simmons, W.R.

1984-01-01

The development of steam generators for the LMFBR was recognized from the onset by the AEC, now DOE, as a difficult, challenging, and high-priority task. The highly reactive nature of sodium with water/steam requires that the sodium-water/steam boundaries of LMFBR steam generators possess a degree of leak-tightness reliability not normally attempted on a commercial scale. In addition, the LMFBR steam generator is subjected to high fluid temperatures and severe thermal transients. These requirements place great demand on materials, fabrication processes, and inspection methods; and even greater demands on the designer to provide steam generators that can meet these demanding requirements, be fabricated without unreasonable shop requirements, and tolerate off-normal effects

Steam Generator Inspection Planning Expert System

International Nuclear Information System (INIS)

Rzasa, P.

1987-01-01

Applying Artificial Intelligence technology to steam generator non-destructive examination (NDE) can help identify high risk locations in steam generators and can aid in preparing technical specification compliant eddy current test (ECT) programs. A steam Generator Inspection Planning Expert System has been developed which can assist NDE or utility personnel in planning ECT programs. This system represents and processes its information using an object oriented declarative knowledge base, heuristic rules, and symbolic information processing, three artificial intelligence based techniques incorporated in the design. The output of the system is an automated generation of ECT programs. Used in an outage inspection, this system significantly reduced planning time

Operating experiences with 1 MW steam generator

Energy Technology Data Exchange (ETDEWEB)

Sano, A; Kanamori, A; Tsuchiya, T

1975-07-01

1 MW steam generator, which was planned as the first stage of steam generator development in Power Reactor and Nuclear Fuel Corp. (PNC) in Japan, is a single-unit, once-through, integrated shell and tube type with multi-helical coil tubes. It was completed in Oarai Engineering Center of PNC in March of 1971, and the various performance tests were carried out up to April, 1972. After the dismantle of the steam generator for structural inspection and material test, it was restored with some improvements. In this second 1 MW steam generator, small leak occurred twice during normal operation. After repairing the failure, the same kind of performance tests as the first steam generator were conducted in order to verify the thermal insulation effect of argon gas in downcomer zone from March to June, 1974. In this paper the above operating experiences were presented including the outline of some performance test results. (author)

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

Response of the steam generator VVER 1000 to a steam line break

International Nuclear Information System (INIS)

Novotny, J.; Novotny, J. Jr.

2003-01-01

Dynamic effects of a steam line break in the weld of the steam pipe and the steam collector on the steam generator system are analyzed. Modelling of a steam line break may concern two cases. The steam line without a restraint and the steam line protected by a whip restraint with viscous elements applied at the postulated break cross-section. The second case is considered. Programme SYSTUS offers a special element the stiffness and viscous damping coefficients of which may be defined as dependent on the relative displacement and velocity of its nodes respectively. A circumferential crack is simulated by a sudden decrease of longitudinal and lateral stiffness coefficients of these special SYSTUS elements to zero. The computation has shown that one can simulate the pipe to behave like completely broken during a time interval of 0,0001 s or less. These elements are used to model the whip restraint with viscous elements and viscous dampers of the GERB type as well. In the case of a whip restraint model the stiffness coefficient-displacement relation and damping coefficient - velocity relation are chosen to fit the given characteristics of the restraint. The special SYSTUS elements are used to constitute Maxwell elements modelling the elasto-plastic and viscous properties of the GERB dampers applied to the steam generator. It has been ascertained that a steam line break at the postulated weld crack between the steam pipe and the steam generator collector cannot endanger the integrity of the system even in a case of the absence of a whip restraint effect. (author)

Removal of secondary sludge from steam generators used in French 900 class nuclear power plants

International Nuclear Information System (INIS)

Lebouc, B.

1982-09-01

The objective is to remove magnetite deposits which have formed on a steam generator tubesheet during plant operation. The deposits are separated from the tubesheet by spraying water at high pressure (about 200 bar at lance nozzle outlets) on each tube bundle ligament, i.e. the spaces between steam generator tubes. The water is recovered in suction lines and then filtered in two seperate units. The residue obtained after settling is removed in the form of solid waste. This paper presents the sludge lancing technique (spray lances, sludge recovery, liquid waste, cooling). A typical operating sequence is detailed (duration, personnel). Specifications for the equipment used are given

US PWR steam generator management: An overview

International Nuclear Information System (INIS)

Welty, C.S. Jr.

1997-01-01

This paper provides an overview on the status of steam generator management activities in US PWRs, and includes: (1) an overview of the impact of steam generator problems; (2) a brief discussion of historical damage trends and the current damage mechanism of most concern; (3) a discussion of the elements of open-quotes steam generator managementclose quotes; and (4) a description of the approach being followed to implement a degradation-specific protocol for tubing inspection and repair. This paper was prepared in conjunction with another paper presented during the Plenary Session of this Conference, open-quotes Steam Generator Degradation: Current Mitigation Strategies for Controlling Corrosionclose quotes, and is provided as a supplement to that material

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

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

The influence of lead on stress corrosion cracking of steam generator tubing

International Nuclear Information System (INIS)

Ryan Curtis Wolfe

2015-01-01

Lead (Pb) is present at low concentrations on the secondary side of steam generators, but is known to accumulate in steam generator sludge and become concentrated in crevices and cracks. Pb is known to have played a role in the degradation of Alloy 600MA tubing, necessitating the replacement of those steam generators. There is new evidence which indicates that Pb has also played a role in the stress corrosion cracking (SCC) of Alloy 600TT. Furthermore. laboratory testing indicates that advanced tubing alloys such as Alloy 690TT and Alloy 800NG area also susceptible to this attack. In response to these vulnerabilities, utilities are attempting to manufacture tubing using processes which will impart optimal corrosion resistance, fabricate and operate SG's to minimize stress in the tubing, undertake efforts to identify and remove the sources of Pb, reduce the existing inventory of Pb using chemical or mechanical cleaning processes, and maintain rigorous chemistry controls. Research is warranted to qualify chemical methods to mitigate PbSCC that may be observed in service. This presentation will review work performed through the Electric Power Research Institute (EPRI) to address the issue of Pb-assisted stress corrosion cracking of steam generator tubing. (author)

Improving electron beam weldability of heavy steel plates for PWR-steam generator

International Nuclear Information System (INIS)

Tomita, Yukio; Mabuchi, Hidesato; Koyama, Kunio

1996-01-01

Installation and replacement of many PWR-steam generators are planned inside and outside Japan. The steel plates for steam generators are heavy in thickness, and increase the number of welding passes and prolong the welding time. Electron beam welding (EBW) can greatly reduce the welding period compared with conventional welding methods (narrow-gap gas metal arc welding (GMAW) and submerged arc welding (SAW)). The problems in applying EBW are to prevent weld defects and to improve the toughness of the weld metal. Defect-free welding procedures were successfully established even in thick steel plates. The factors that deteriorate weld-metal (WM) toughness of EBW were investigated. The manufacturing process, which utilizes a new secondary refining process at steelmaking and a high-torque mill at plate mill in actual mass-production, were established. EBW base metal and WM have better properties including fracture toughness than those of conventional welding processes. As a result, an application of EBW to the fabrication of PWR-steam generators has become possible. Large amounts of ASTM A533 Gr B Cl 2 (JIS SQV2B) steel plates in actual PWR-steam generators have come to be produced (more than 1,500 ton) by applying EBW. (author)

Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

International Nuclear Information System (INIS)

Mendler, O.J.; Takeuchi, K.; Young, M.Y.

1986-10-01

The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results

Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

Energy Technology Data Exchange (ETDEWEB)

Mendler, O J; Takeuchi, K; Young, M Y

1986-10-01

The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.

Design of PFBR steam generator

International Nuclear Information System (INIS)

Chetal, S.C.; Bhoje, S.B.; Mitra, T.K.; Paranjpe, S.R.; Vaidyanathan, G.

1990-01-01

Vertical straight tube with an expansion bend in sodium path is the design selected for the steam generators of 500 MWe Prototype Fast Breeder Reactor (PFBR). There are 4 secondary loops with each loop consisting of 3 modules. With sodium reheat incorporated each module comprises of one evaporator, superheater and reheater. Material of construction is 2.25Cr-1Mo for evaporator and 9Cr-1Mo for superheater and reheater. The tube to tubesheet weld is internal bore butt weld with tubesheet having raised spigot. Aim is to have reliable design with higher plant availability. Design considerations leading to the choice of design features selected are discussed in the paper and a ''reference'' design has been described. (author). 2 figs, 1 tab

High temperature technological heat exchangers and steam generators with helical coil assembly tube bundle

International Nuclear Information System (INIS)

Korotaev, O.J.; Mizonov, N.V.; Nikolaevsky, V.B.; Nazarov, E.K.

1990-01-01

Analysis of thermal hydraulics characteristics of nuclear steam generators with different tube bundle arrangements and waste heat boilers for ammonia production units was performed on the basis of operating experience results and research and development data. The present report involves the obtained information. The estimations of steam generator performances and repair-ability are given. The significant temperature profile of the primary and secondary coolant flows are attributed to all steam generator designs. The intermediate mixing is found to be an effective means of temperature profile overcoming. At present the only means to provide an effective mixing in heat exchangers of the following types: straight tubes, field tubes, platen tubes and multibank helical coil tubes (with complicated bend distribution along their length) are section arrangements in series in conjunction with forced and natural mixing in connecting lines. Development of the unificated system from mini helical coil assemblies allows to design and manufacture heat exchangers and steam generators within the wide range of operating conditions without additional expenses on the research and development work

Duke Power Company - McGuire Nuclear Station: steam-generator hideout return and cleanup

International Nuclear Information System (INIS)

Anon.

1986-01-01

McGuire Nuclear Station steam generator hideout return and cleanup are discussed. Hideout return data are presented for Unit 1 shutdowns that occurred on November 23, 1984, and April 19, 1985, and a Unit 2 shutdown on January 25, 1985. The data are presented as the concentrations of various species as a function of time after power reduction and primary water temperature. The steam generator blowdown as a function of time after power reduction is also presented. The concentrations of sodium, potassium, calcium, magnesium, aluminum, iron, and copper cations, and chloride, fluoride, sulfate, phosphate and nitrite anions were monitored during the each shutdown. Silica was also measured in the two 1985 shutdowns. The return of sulfate, phosphate, calcium and magnesium showed retrograde solubility. Silica concentrations showed an increase as the temperature decreased to about 450 to 500 0 F and then they decreased as the temperature decreased. McGuire has a holf point at 300 at 350 0 F to clean up the steam generator secondary water. The return of sulfates should occur within 4 to 6 hours. The blowdown is maximized to reduce the secondary water impurity concentrations. Cleanup continues until the sulfate concentration is reduced to below 100 ppb. At that point cooldown is continued

Dynamic model of YGN 3 and 4 steam generators for natural circulation mode

International Nuclear Information System (INIS)

Sohn, Jong Joo

1995-02-01

A dynamic model for the secondary side of Yonggwang nuclear power plant units 3 and 4 (YGN 3 and 4) steam generator model is developed to improve the accuracy of the present performance analysis code. The new model is based on the one-dimensional three region model to predict the local quality and void fraction distribution along the U-tube length. The local quality concept is used instead of the Wilson bubble rise correlation to simulate the steam generators in the natural circulation mode. The new model can be applicable to the plants in the power operation modes such as load maneuvering transients in which the steam generator internal flow is maintained in the natural circulation mode. To validate the new model, the code predictions are compared with the actual plant data measured for the selected load maneuvering tests performed during the YGN units 3 power ascension test period. The results from the improved model show better agreement with the plant data than those from the present code. Especially, the new model's capability of accurately simulating the steam generator water level behavior during the fast transients such as a large load rejection event is demonstrated

Future steam generator designs. Single wall designs

International Nuclear Information System (INIS)

Hayden, O.

1978-01-01

The easily removable 'U' tube design style adopted in the UK for the existing PFR Steam Generators, the Replacement Units now in production and for the future CDFR, gives the operator an extremely valuable option in the event of a water/steam leak occurring inside the Steam Generator. He can choose to shut-down, attempt to find the leak, assess damage, repair, revalidate and return to service in situ, or he can elect to remove the defect unit and replace with a 'proven' spare before returning the circuit to power. With the latter approach the resultant outage time is a known entity of about two weeks. If a repair is attempted in situ, predictions of outage time can become a matter of guesswork since one has no 'guaranteed' method of leak location and the assessment of secondary damage may be very time consuming, depending on the size and type of the original leak together with the particular design style of the Steam Generator. A further significant advantage of the removable 'U' tube design concept is that periodic interchanging of bundles with a spare enables routine chemical cleaning and thorough scheduled tube inspections, with specimen tube sample removal if required for monitoring purposes. If necessary, bundle decontamination can be undertaken to assess engineering deterioration to various degrees of thoroughness ranging from 100% equivalent factory final assembly inspection, to partial decontamination operating via a glovebox type of maintenance bag arrangement, examining local points of both shell and tube areas of the bundle. Many lessons from the last five years' experience of PFR will be incorporated into the design of the CDFR and PFR Steam Generators have two very good examples of how the designer can ease or severely handicap the operator in coping with sodium/water leakages. Good, quick access to tube ends is achieved in the existing PFR Evaporator by simply unbolting the steam/water closure head, but on the superheater and reheater hand-caps have

Future steam generator designs. Single wall designs

Energy Technology Data Exchange (ETDEWEB)

Hayden, O [Nuclear Power Company Ltd, Warrington, Cheshire (United Kingdom)

1978-10-01

The easily removable 'U' tube design style adopted in the UK for the existing PFR Steam Generators, the Replacement Units now in production and for the future CDFR, gives the operator an extremely valuable option in the event of a water/steam leak occurring inside the Steam Generator. He can choose to shut-down, attempt to find the leak, assess damage, repair, revalidate and return to service in situ, or he can elect to remove the defect unit and replace with a 'proven' spare before returning the circuit to power. With the latter approach the resultant outage time is a known entity of about two weeks. If a repair is attempted in situ, predictions of outage time can become a matter of guesswork since one has no 'guaranteed' method of leak location and the assessment of secondary damage may be very time consuming, depending on the size and type of the original leak together with the particular design style of the Steam Generator. A further significant advantage of the removable 'U' tube design concept is that periodic interchanging of bundles with a spare enables routine chemical cleaning and thorough scheduled tube inspections, with specimen tube sample removal if required for monitoring purposes. If necessary, bundle decontamination can be undertaken to assess engineering deterioration to various degrees of thoroughness ranging from 100% equivalent factory final assembly inspection, to partial decontamination operating via a glovebox type of maintenance bag arrangement, examining local points of both shell and tube areas of the bundle. Many lessons from the last five years' experience of PFR will be incorporated into the design of the CDFR and PFR Steam Generators have two very good examples of how the designer can ease or severely handicap the operator in coping with sodium/water leakages. Good, quick access to tube ends is achieved in the existing PFR Evaporator by simply unbolting the steam/water closure head, but on the superheater and reheater hand-caps have

Parametric study for horizontal steam generator modelling

Energy Technology Data Exchange (ETDEWEB)

Ovtcharova, I. [Energoproekt, Sofia (Bulgaria)

1995-12-31

In the presentation some of the calculated results of horizontal steam generator PGV - 440 modelling with RELAP5/Mod3 are described. Two nodalization schemes have been used with different components in the steam dome. A study of parameters variation on the steam generator work and calculated results is made in cases with separator and branch.

Parametric study for horizontal steam generator modelling

Energy Technology Data Exchange (ETDEWEB)

Ovtcharova, I [Energoproekt, Sofia (Bulgaria)

1996-12-31

In the presentation some of the calculated results of horizontal steam generator PGV - 440 modelling with RELAP5/Mod3 are described. Two nodalization schemes have been used with different components in the steam dome. A study of parameters variation on the steam generator work and calculated results is made in cases with separator and branch.

Steam generator arrangement

International Nuclear Information System (INIS)

Ssinegurski, E.

1981-01-01

A steam flow path arrangement for covering the walls of the rear gas pass of a steam generator is disclosed. The entire flow passes down the sidewalls with a minor portion then passing up through the rear wall to a superheater inlet header at an intermediate elevation. The major portion of the flow passes up the front wall and through hanger tubes to a roof header. From there the major portion passes across the roof and down the rear wall to the superheater inlet header at the intermediate elevation

CRBRP steam-generator design evolution

International Nuclear Information System (INIS)

Geiger, W.R.; Gillett, J.E.; Lagally, H.O.

1983-01-01

The overall design of the CRBRP Steam Generator is briefly discussed. Two areas of particular concern are highlighted and considerations leading to the final design are detailed. Differential thermal expansion between the shell and the steam tubes is accommodated by the tubes flexing in the curved section of the shell. Support of the tubes by the internals structure is essential to permit free movement and minimize tube wear. Special spacer plate attachment and tube hole geometry promote unimpeded axial movement of the tubes by allowing individual tubes to rotate laterally and by providing lateral movement of the spacer plates relative to the adjacent support structure. The water/steam heads of the CRBRP Steam Generator are spherical heads welded to the lower and upper tubesheets. They were chosen principally because they provide a positively sealed system and result in more favorable stresses in the tubesheets when compared to mechanically attached steamheads

Wavelet network controller for nuclear steam generators

International Nuclear Information System (INIS)

Habibiyan, H; Sayadian, A; Ghafoori-Fard, H

2005-01-01

Poor control of steam generator water level is the main cause of unexpected shutdowns in nuclear power plants. Particularly at low powers, it is a difficult task due to shrink and swell phenomena and flow measurement errors. In addition, the steam generator is a highly complex, nonlinear and time-varying system and its parameters vary with operating conditions. Therefore, it seems that design of a suitable controller is a necessary step to enhance plant availability factor. The purpose of this paper is to design, analyze and evaluate a water level controller for U-tube steam generators using wavelet neural networks. Computer simulations show that the proposed controller improves transient response of steam generator water level and demonstrate its superiority to existing controllers

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

Vibration Analysis for Steam Dryer of APR1400 Steam Generator

Energy Technology Data Exchange (ETDEWEB)

Han, Sung-heum; Ko, Doyoung [KHNP CRI, Daejeon (Korea, Republic of); Cho, Minki [Doosan Heavy Industry, Changwon (Korea, Republic of)

2016-10-15

This paper is related to comprehensive vibration assessment program for APR1400 steam generator internals. According to U.S. Nuclear Regulatory Commission, Regulatory Guide 1.20 (Rev.3, March 2007), we conducted vibration analysis for a steam dryer as the second steam separator of steam generator internals. The vibration analysis was performed at the 100 % power operating condition as the normal operation condition. The random hydraulic loads were calculated by the computational fluid dynamics and the structural responses were predicted by power spectral density analysis for the probabilistic method. In order to meet the recently revised U.S. NRC RG 1.20 Rev.3, the CVAP against the potential adverse flow effects in APR1400 SG internals should be performed. This study conducted the vibration response analysis for the SG steam dryer as the second moisture separator at the 100% power condition, and evaluated the structural integrity. The predicted alternating stress intensities were evaluated to have more than 17.78 times fatigue margin compared to the endurance limit.

Evaluation of materials' corrosion and chemistry issues for advanced gas cooled reactor steam generators using full scale plant simulations

International Nuclear Information System (INIS)

Woolsey, I.S.; Rudge, A.J.; Vincent, D.J.

1998-01-01

Advanced Gas Cooled Reactors (AGRS) employ once-through steam Generators of unique design to provide steam at approximately 530 degrees C and 155 bar to steam turbines of similar design to those of fossil plants. The steam generators are highly compact, and have either a serpentine or helical tube geometry. The tubes are heated on the outside by hot C0 2 gas, and steam is generated on the inside of the tubes. Each individual steam generator tube consists of a carbon steel feed and primary economiser section, a 9%Cr steel secondary economiser, evaporator and primary superheater, and a Type 316L austenitic stainless steel secondary superheater, all within a single tube pass. The multi-material nature of the individual tube passes, the need to maintain specific thermohydraulic conditions within the different material sections, and the difficulties of steam generator inspection and repair, have required extensive corrosion-chemistry test programmes to ensure waterside corrosion does not present a challenge to their integrity. A major part of these programmes has been the use of a full scale steam generator test facility capable of simulating all aspects of the waterside conditions which exist in the plant. This facility has been used to address a wide variety of possible plant drainage/degradation processes. These include; single- and two-phase flow accelerated corrosion of carbon steel, superheat margins requirements and the stress-corrosion behaviour of the austenitic superheaters, on-load corrosion of the evaporator materials, and iron transport and oxide deposition behaviour. The paper outlines a number of these, and indicates how they have been of value in helping to maintain reliable operation of the plant. (author)

Dynamic modelling of nuclear steam generators

International Nuclear Information System (INIS)

Kerlin, T.W.; Katz, E.M.; Freels, J.; Thakkar, J.

1980-01-01

Moving boundary, nodal models with dynamic energy balances, dynamic mass balances, quasi-static momentum balances, and an equivalent single channel approach have been developed for steam generators used in nuclear power plants. The model for the U-tube recirculation type steam generator is described and comparisons are made of responses from models of different complexity; non-linear versus linear, high-order versus low order, detailed modeling of the control system versus a simple control assumption. The results of dynamic tests on nuclear power systems show that when this steam generator model is included in a system simulation there is good agreement with actual plant performance. (author)

Future development LMFBR-steam generators SNR2

International Nuclear Information System (INIS)

Essebaggers, J.; Pors, J.G.

1975-01-01

The development work for steam generators for large LMFBR plants by Neratoom will be reviewed consisting of: 1. Development engineering information. 2. Concept select studies followed by conceptual designs of selected models. 3. Development manufacturing techniques. 4. Detail design of a prototype unit. 5. Testing of sub-constructions for prototype steam generators. In this presentation item 1 and 2 above will be high lighted, identifying the development work for the SNR-2 steam generators on short term basis. (author)

Laboratory results of stress corrosion cracking of steam generator tubes in a complex environment - An update

Energy Technology Data Exchange (ETDEWEB)

Horner, Olivier; Pavageau, Ellen-Mary; Vaillant, Francois [EDF R and D, Materials and Mechanics of Components Department, 77818 Moret-sur-Loing (France); Bouvier, Odile de [EDF Nuclear Engineering Division, Centre d' Expertise et d' Inspection dans les Domaines de la Realisation et de l' Exploitation, 93206 Saint Denis (France)

2004-07-01

Stress corrosion cracking occurs in the flow-restricted areas on the secondary side of steam generator tubes of Pressured Water Reactors (PWR), where water pollutants are likely to concentrate. Chemical analyses carried out during the shutdowns gave some insight into the chemical composition of these areas, which has evolved during these last years (i.e. less sodium as pollutants). It has been modeled in laboratory by tests in two different typical environments: the sodium hydroxide and the sulfate environments. These models satisfactorily describe the secondary side corrosion of steam generator tubes for old plant units. Furthermore, a third typical environment - the complex environment - which corresponds to an All Volatile Treatment (AVT) environment containing alumina, silica, phosphate and acetic acid has been recently studied. This particular environment satisfactorily reproduces the composition of the deposits observed on the surface of the steam generator tubes as well as the degradation of the tubes. A review of the recent laboratory results obtained by considering the complex environment are presented here. Several tests have been carried out in order to study initiation and propagation of secondary side corrosion cracking for some selected materials in such an environment. 600 Thermally Treated (TT) alloy reveals to be less sensitive to secondary side corrosion cracking than 600 Mill Annealed (MA) alloy. Finally, the influence of some related factors like stress, temperature and environmental factors are discussed. (authors)

Laboratory results of stress corrosion cracking of steam generator tubes in a complex environment - An update

International Nuclear Information System (INIS)

Horner, Olivier; Pavageau, Ellen-Mary; Vaillant, Francois; Bouvier, Odile de

2004-01-01

Stress corrosion cracking occurs in the flow-restricted areas on the secondary side of steam generator tubes of Pressured Water Reactors (PWR), where water pollutants are likely to concentrate. Chemical analyses carried out during the shutdowns gave some insight into the chemical composition of these areas, which has evolved during these last years (i.e. less sodium as pollutants). It has been modeled in laboratory by tests in two different typical environments: the sodium hydroxide and the sulfate environments. These models satisfactorily describe the secondary side corrosion of steam generator tubes for old plant units. Furthermore, a third typical environment - the complex environment - which corresponds to an All Volatile Treatment (AVT) environment containing alumina, silica, phosphate and acetic acid has been recently studied. This particular environment satisfactorily reproduces the composition of the deposits observed on the surface of the steam generator tubes as well as the degradation of the tubes. A review of the recent laboratory results obtained by considering the complex environment are presented here. Several tests have been carried out in order to study initiation and propagation of secondary side corrosion cracking for some selected materials in such an environment. 600 Thermally Treated (TT) alloy reveals to be less sensitive to secondary side corrosion cracking than 600 Mill Annealed (MA) alloy. Finally, the influence of some related factors like stress, temperature and environmental factors are discussed. (authors)

WWER Steam Generators Tubing Performance and Aging Management

International Nuclear Information System (INIS)

Trunov, Nikolay B.; Davidenko, Stanislav E.; Grigoriev, Vladimir A.; Popadchuk, Valery S.; Brykov, Sergery I.; Karzov, Georgy P.

2006-01-01

At WWER NPPs the horizontal steam generators (SGs), are used that differ in design concept from vertical SGs mostly used at western NPPs. Reliable operation of SG heat-exchanging tubes is the crucial worldwide problem for NPP of various types. According to the operation feedback the water chemistry is the governing factor affecting operability of SG tubing. The secondary side corrosion is considered to be the main mechanism of SG heat-exchanging tubes damage at WWER plants. To make the assessment of the tubing integrity the combination of pressure tests and eddy-current tests is used. Assessment of the tubing performance is an important part of SG life extension practice. The given paper deals with the description of the tube testing strategy and the approach to tube integrity assessment based on deterministic and probabilistic methods of fracture mechanics. Requirements for eddy-current test are given as well. Practice of condition monitoring and implementing the database on steam generators operation are presented. The approach to tubes plugging criteria is described. The research activities on corrosion mechanism studies and residual lifetime evaluation are mentioned. (authors)

Steam generator with U-tube bank arranged within an oblong pressure vessel

International Nuclear Information System (INIS)

Beckmann, G.; Fritz, K.

1976-01-01

This steam generator equipped with a U-tube bundle differs substantially from standard types because of its operational condition. The boiler is at a tilt of 45 0 , the piping base, the inlet, and the outlet for the primary medium are arranged at the top. This improves the heat flow of the secondary medium within the boiler. The steam room placed near the piping base is enlarged on the hot side of the U-tube bundle due to the tilt of the water level, allowing drying and overheating of the steam without additional mounting of water separators and special overheaters. The additional space obtained by this construction is estimated at 6%. (FW) [de

French steam generator

International Nuclear Information System (INIS)

Remond, A.

1986-01-01

After recalling the potential damage mode of tubes of steam generator, the author recalls the safety criteria used in France. The improvements and the process of damage prejudice and reparation for tubular bundle are presented [fr

Model studies of the vertical steam generator thermal-hydraulic characteristics

International Nuclear Information System (INIS)

Desyatun, V.F.; Moskvichev, V.F.; Ulasov, V.M.; Morozov, V.G.; Burkov, V.K.; Grebennikov, V.N.

1984-01-01

Results of investigations conducted to clarify the calculation technique and to test the workability of the main elements and units of the PGV-250 vertical steam generator of saturated steam are considered. The steam generating capacity of the plant is 1486 t/h, thermal power is 792 MW. Steam generation follows a multiple circulation scheme. The heat surface comprises 330-shields. The investigations are carried out with a model which reproduces all the main elements of the steam generator xcluding the economizer section. The flow rates of feed water, generated steam and coolant of the first circuit as well as temperature, pressure and humidity of the generated steam past the separator are determined. The average heat transfer factors of the heat surface are calculated on the base of the data obtained and a conclusion is drawn on the correctness of the thermohydraulic calculation technique used in development of the PGV-250 steam generator design. Temperature pulsations and heat surface steaming are not observed. The steam humidity at the outlet and steam capture into sinking tubes are within permissible values

LMR steam generator blowdown with RETRAN

International Nuclear Information System (INIS)

Wei, T.Y.C.

1985-01-01

One of the transients being considered in the FSAR Chapter 15 analyses of anticipated LMR transients is the fast blowdown of a steam generator upon inadvertent actuation of the liquid metal/water reaction mitigation system. For the blowdown analysis, a stand-alone steam generator model for the IFR plant was constructed using RETRAN

Evaluation of EDTA based chemical formulations for the cleaning of monel-400 tubed steam generators

International Nuclear Information System (INIS)

Velmurugan, S.; Rufus, A.L.; Sathyaseelan, V.S.; Kumar, P.S.; Veena, S.N.; Srinivasan, M.P.; Narasimhan, S.V.

1998-01-01

The Steam Generator (SG) is an important component in any nuclear power plant which contributes significantly for the over all performance of the reactor. The failure of SG tubes occurs mainly by corrosion under accelerated conditions caused by fouling. There is continuous ingress of the corrosion products and ionic impurities from the condenser and feed train of the secondary heat transfer system. The corrosion products accumulate in the stagnant areas near the tube sheet, over the tube support plates and in the tube to tube support plate crevices. These accumulated deposits help to concentrate the aggressive impurities and induce a variety of corrosion processes affecting the structural materials and finally leading to failure of the SG tube. Scale forming impurities can deposit over the tube surfaces and result in reduction of heat transfer efficiency and over heating of the surfaces. Every effort is being made to control the transport of impurities to the steam generator. Increased blow down, installation of condensate polishers and use of all volatile amines have helped to reduce the corrosion product and ionic impurities input into the steam generators of PHWRs. Despite these efforts, failures of SG tubes in PHWRs have been reported. Hence, attempts are being made to develop chemical formulations to clean the deposits accumulated in the steam generators. The EPRI-SGOG chemical cleaning process has been tried with good success in steam generators of different designs including the steam generators of PHWRs. This paper discusses the work on the evaluation of EDTA based chemical cleaning formulations for monel-400 tubed steam generators of PHWRs. (author)

Preventing degradation and in-service inspection of NPP steam generator

International Nuclear Information System (INIS)

Ding Xunshen

1999-01-01

The author describes the degradation of steam generator tubes in initial operating stage of France NPP. The author emphatically presents the preventive measures to tackle degradation and In-service inspection of Daya Bay NPP 1 unit as a guarantee of safety operation, including secondary side water chemistry monitoring and controlling, leakage rate monitoring, eddy-current inspection, mechanical cleaning, cleanliness inspection, foreign objects removal and etc

Research program plan: steam generators

International Nuclear Information System (INIS)

Muscara, J.; Serpan, C.Z. Jr.

1985-07-01

This document presents a plan for research in Steam Generators to be performed by the Materials Engineering Branch, MEBR, Division of Engineering Technology, (EDET), Office of Nuclear Regulatory Research. It is one of four plans describing the ongoing research in the corresponding areas of MEBR activity. In order to answer the questions posed, the Steam Generator Program has been organized with the three elements of non-destructive examination; mechanical integrity testing; and corrosion, cleaning and decontamination

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

Identification of leaky steam generators by iodine mapping technique and development of tools for cutting of tubes of steam generators of Indian PHWRS

International Nuclear Information System (INIS)

Subba Rao, D.

2006-01-01

inspected in previous ISI and no reportable indications were observed. To investigate the cause of steam generator tubes leak two failed tubes were cut and removed for failure analysis. To perform this activity some special tools were designed and developed in house and whole job of two failed tubes cutting, removal and plugging with specially developed extended plugs for left out portion of the cut tubes support was executed with in four days. After removing failed tubes, one S.S metallic gasket strip (foreign material) was found stuck between two failed tubes and same was removed using special tools. Based on metallurgical and chemical analysis the root cause for tubes failure was due to fretting action by foreign material inclusion, i.e. a metallic strip. A video scope was taken to assess the structural integrity of internals of primary and secondary side of the steam generator and it was found okay. Both S.S gasket metallic strip and failed tubes were tested for metallurgical analysis for hardness and found that the gasket strip harder than SG tube material. Feed water control valves maintenance procedures were revised and all the maintenance personnel were trained and familiarized to prevent the broken gasket pieces entering in to Steam generators through feed water. Based on metallurgical and chemical analysis the Steam generator tubes are healthy. (author)

Evaluation of hideout return data from U.S. PWR steam generators

International Nuclear Information System (INIS)

Connor, W.M.; Richards, J.E.

1988-01-01

Since the middle to late 1970's, dramatic reductions in the quantities of impurities in the bulkwater of PWR steam generators have been made by U.S. utilities. Today most utilities operate at full power with impurity concentrations in the steam generator blowdown in the low ppb range, well within existing industry guideline control limits. Despite these efforts, some of these same utilities have subsequently encountered secondary side stress corrosion cracking (SCC) and intergranular attack (IGA) of steam generator tubing within deep tubesheet crevices and more recently at tube support intersections. It must, therefore, be concluded that either continuous low level input of contaminants within existing guideline limits, or intermittent short duration input, undetected by either current sampling and analysis techniques or procedures, are permitting ingress of corrosive impurity species which subsequently concentrate in flow-occluded regions to produce localized tube corrosion. To better understand both the quantity and composition of accumulated impurity species, more and more utilities, even those who have not experienced any steam generator corrosion, have begun to perform rigorous sampling and analysis evaluations of returning chemical contaminants each time the units are brought off-line. This paper will show examples of how these data are being used by U.S. industry to gain valuable information about accumulated contaminant inventories, to make cycle-to-cycle and plant-to-plant comparisons, and to develop plant specific actions to promote maximum contaminant removal. (author)

Particle deposition modeling in the secondary side of a steam generator bundle model

Energy Technology Data Exchange (ETDEWEB)

Mukin, Roman, E-mail: roman.mukin@psi.ch; Dehbi, Abdel, E-mail: abdel.dehbi@psi.ch

2016-04-01

A steam generator (SG) tube rupture (SGTR) model is studied in this paper. This model based on a experimental facility called Aerosol Trapping In a Steam Generator (ARTIST), which is a model of a scaled steam generator tube bundle consisting of 270 tubes and a guillotine tube to address aerosol deposition phenomena on two different scales: near the tube break, where the gas velocities and turbulence are very intensive, and far away from the break, where the flow velocities are three orders of magnitude lower. Owing to complexity of the flow, 3D simulations with highly resolved computational mesh near the break were done. First, the flow inside an isolated tube with a guillotine tube break has been studied in the framework of Reynolds Averaged Navier Stokes (RANS) approach. The next part is devoted to the simulation of an inclined gas jet entering the SG tube bundle via the guillotine tube breach with more advanced CFD tools. In particular, Detached Eddy Simulation (DES) and RANS are applied to tackle the wide range of flow scales. Flow field velocity comparison showed that DES results are reproducing wavy structure of the flow field in far field from the break observed in experiment. Particle transport and deposition is modelled by Lagrangian continuous random walk (CRW) model, which has been developed and validated previously. It is found that the DES combined with the CRW to supply fluctuating velocity components predicts deposition rates that are generally within the scatter of the measured data. Monodisperse, spherical SiO{sub 2} particles with AMMD = 1.4 μm were used as aerosol particles in simulations. To be economically feasible, the computations were made with the open source CFD code OpenFOAM. Comparison of the calculated flow with the experimental axial velocity distribution data at different vertical levels has been performed.

Mathematical modeling of control system for the experimental steam generator

Science.gov (United States)

Podlasek, Szymon; Lalik, Krzysztof; Filipowicz, Mariusz; Sornek, Krzysztof; Kupski, Robert; Raś, Anita

2016-03-01

A steam generator is an essential unit of each cogeneration system using steam machines. Currently one of the cheapest ways of the steam generation can be application of old steam generators came from army surplus store. They have relatively simple construction and in case of not so exploited units - quite good general conditions, and functionality of mechanical components. By contrast, electrical components and control systems (mostly based on relay automatics) are definitely obsolete. It is not possible to use such units with cooperation of steam bus or with steam engines. In particular, there is no possibility for automatically adjustment of the pressure and the temperature of the generated steam supplying steam engines. Such adjustment is necessary in case of variation of a generator load. The paper is devoted to description of improvement of an exemplary unit together with construction of the measurement-control system based on a PLC. The aim was to enable for communication between the steam generator and controllers of the steam bus and steam engines in order to construction of a complete, fully autonomic and maintenance-free microcogeneration system.

Mitigation of organically bound sulphate from water treatment plants at Bruce NGS and impact on steam generator secondary side chemistry control

Energy Technology Data Exchange (ETDEWEB)

Nashiem, R.; Davloor, R.; Harper, B.; Smith, K. [Bruce Power, Tiverton, Ontario (Canada); Gauthier, C. [CTGIX Services Inc., Burlington, Ontario (Canada); Schexnailder, S. [GE Water and Process Technologies, Dallas, Texas (United States)

2010-07-01

Bruce Power is the source of more than 20 per cent of Ontario's electricity and currently operates six reactor units at the Bruce Nuclear Generating Station A (two units) and B (four units) stations located on Lake Huron. This paper discusses the challenges faced and operating experience (OPEX) gained in meeting WANO 1.0 chemistry performance objectives for steam generator secondary side chemistry control, particularly with control of steam generator sulphates. A detailed sampling and analysis program conducted as part of this study concluded that a major contributor to steam generator (SG) elevated sulphates is Organically Bound Sulphate (OBS) in Water Treatment Plants (WTP) effluent. The Bruce A and B WTPs consist of clarification with downstream sand and carbon filtration for Lake Water pre-treatment, which are followed by conventional Ion Exchange (IX) demineralization. Samples taken from various locations in the process stream were analyzed for a variety of parameters including both organic bound and inorganic forms of sulphate. The results are inconclusive with respect to finding the definitive source of OBS. This is primarily due to the condition that the OBS in the samples, which are in relatively low levels, are masked during chemical analysis by the considerably higher inorganic sulphate background. Additionally, it was also determined that on-line Total Organic Carbon (TOC) levels at different WTP locations did not always correlate well with OBS levels in the effluent, such that TOC could not be effectively used as a control parameter to improve OBS performance of the WTP operation. Improvement efforts at both plants focused on a number of areas including optimization of clarifier operation, replacement of IX resins, addition of downstream mobile polishing trailers, testing of new resins and adsorbents, pilot-scale testing with a Reverse Osmosis (RO) rig, review of resin regeneration and backwashing practices, and operating procedure improvements

Mitigation of organically bound sulphate from water treatment plants at Bruce NGS and impact on steam generator secondary side chemistry control

International Nuclear Information System (INIS)

Nashiem, R.; Davloor, R.; Harper, B.; Smith, K.; Gauthier, C.; Schexnailder, S.

2010-01-01

Bruce Power is the source of more than 20 per cent of Ontario's electricity and currently operates six reactor units at the Bruce Nuclear Generating Station A (two units) and B (four units) stations located on Lake Huron. This paper discusses the challenges faced and operating experience (OPEX) gained in meeting WANO 1.0 chemistry performance objectives for steam generator secondary side chemistry control, particularly with control of steam generator sulphates. A detailed sampling and analysis program conducted as part of this study concluded that a major contributor to steam generator (SG) elevated sulphates is Organically Bound Sulphate (OBS) in Water Treatment Plants (WTP) effluent. The Bruce A and B WTPs consist of clarification with downstream sand and carbon filtration for Lake Water pre-treatment, which are followed by conventional Ion Exchange (IX) demineralization. Samples taken from various locations in the process stream were analyzed for a variety of parameters including both organic bound and inorganic forms of sulphate. The results are inconclusive with respect to finding the definitive source of OBS. This is primarily due to the condition that the OBS in the samples, which are in relatively low levels, are masked during chemical analysis by the considerably higher inorganic sulphate background. Additionally, it was also determined that on-line Total Organic Carbon (TOC) levels at different WTP locations did not always correlate well with OBS levels in the effluent, such that TOC could not be effectively used as a control parameter to improve OBS performance of the WTP operation. Improvement efforts at both plants focused on a number of areas including optimization of clarifier operation, replacement of IX resins, addition of downstream mobile polishing trailers, testing of new resins and adsorbents, pilot-scale testing with a Reverse Osmosis (RO) rig, review of resin regeneration and backwashing practices, and operating procedure improvements

RELAP5/MOD2 code modifications to obtain better predictions for the once-through steam generator

International Nuclear Information System (INIS)

Blanchat, T.; Hassan, Y.

1989-01-01

The steam generator is a major component in pressurized water reactors. Predicting the response of a steam generator during both steady-state and transient conditions is essential in studying the thermal-hydraulic behavior of a nuclear reactor coolant system. Therefore, many analytical and experimental efforts have been performed to investigate the thermal-hydraulic behavior of the steam generators during operational and accident transients. The objective of this study is to predict the behavior of the secondary side of the once-through steam generator (OTSG) using the RELAP5/MOD2 computer code. Steady-state conditions were predicted with the current version of the RELAP5/MOD2 code and compared with experimental plant data. The code predictions consistently underpredict the degree of superheat. A new interface friction model has been implemented in a modified version of RELAP5/MOD2. This modification, along with changes to the flow regime transition criteria and the heat transfer correlations, correctly predicts the degree of superheat and matches plant data

Steam generator tube fitness-for-service guidelines

International Nuclear Information System (INIS)

Gorman, J.A.; Harris, J.E.; Lowenstein, D.B.

1995-07-01

The objectives of this project were to characterize defect mechanisms which could affect the integrity of steam generator tubes, to review and critique state-of-the-art Canadian and international steam generator tube fitness-for-service criteria and guidelines, and to obtain recommendations for criteria that could be used to assess fitness-for service guidelines for steam generator tubes containing defects in Canadian power plant service. Degradation mechanisms, that could affect CANDU steam generator tubes in Canada, have been characterized. The design standards and safety criteria that apply to steam generator tubing in nuclear power plant service in Canada and in Belgium, France, Japan, Spain, Sweden, and the USA have been reviewed and described. The fitness-for-service guidelines used for a variety of specific defect types in Canada and internationally have been evaluated and described in detail in order to highlight the considerations involved in developing such defect specific guidelines. Existing procedures for defect assessment and disposition have been identified, including inspection and examination practices. The approaches used in Canada and in Belgium, France, Japan, Spain, Sweden, and the USA for fitness-for-service guidelines were compared and contrasted for a variety of defect mechanisms. The strengths and weaknesses of the various approaches have been assessed. The report presents recommendations on approaches that may be adopted in the development of fitness-for-service guidelines for use in the dispositioning of steam generator tubing defects in Canada. (author). 175 refs., 2 tabs., 28 figs

Secondary side corrosion in steam generator tubes: lessons learned in France from the in-service inspection results

International Nuclear Information System (INIS)

Comby, R.

1997-01-01

Non-destructive testing (NDT) has proved to be very important in the maintenance of steam generator tubing. This is particularly true in the case of secondary side corrosion, because this type of degradation leads to various morphologies which are often complex (intergranular attack) (IGA), intergranular stress corrosion cracking (IGSCC), or a mixture of both. Their detection and characterization by the usual NDT techniques have been achieved through numerous laboratory studies, which were conducted in order to determine the performance and limitations of NDT. Pulled tube examination in a hot laboratory was very valuable, for both NDT and fracture mechanics aspects. The eddy current bobbin coil probe, used for multipurpose inspection of tubes, allows the detection of IGA-SCC at the tube support plate elevation. In France, the use of rotating probes is not required for that type of degradation, since the repair criterion is based on bobbin coil results only. The bobbin coil is also used for detection of IGSCC occurring in free spans, within sludge deposits. The eddy current rotating probe allows, in that case, characterization of main cracks. Concerning the outer diameter initiated circumferential cracks which occur at the top of the tube sheet, only the rotating probe is used. An ultrasonic (UT) inspection was performed several times, in order to obtain information on UT capabilities. The goal of tube inspection is obviously knowledge of the status of steam generators, but also to follow up degradations and to estimate their revolution, and to verify the beneficial effect of some corrective measures, e.g. boric acid injection. (orig.)

Wastage of Steam Generator Tubes by Sodium-Water Reaction

International Nuclear Information System (INIS)

Jeong, Ji Young; Kim, Jong Man; Kim, Tae Joon; Choi, Jong Hyeun; Kim, Byung Ho; Lee, Yong Bum; Park, Nam Cook

2010-01-01

The Korea Advanced LIquid MEtal Reactor (KALIMER) steam generator is a helical coil, vertically oriented, shell-and-tube type heat exchanger with fixed tube-sheet. The conceptual design and outline drawing of the steam generator are shown in Figure 1. Flow is counter-current, with sodium on the shell side and water/steam on the tube side. Sodium flow enters the steam generator through the upper inlet nozzles and then flows down through the tube bundle. Feedwater enters the steam generator through the feedwater nozzles at the bottom of steam generator. Therefore, if there is a hole or a crack in a heat transfer tube, a leakage of water/steam into the sodium may occur, resulting in a sodium-water reaction. When such a leak occurs, so-called 'wastage' is the result which may cause damage to or a failure of the adjacent tubes. If a steam generator is operated for some time in this condition, it is possible that it might create an intermediate leak state which would then give rise to the problems of a multi-target wastage in a very short time. Therefore, it is very important to predict these phenomena quantitatively from the view of designing a steam generator and its leak detection systems. For this, multi-target wastage tests for modified 9Cr-1Mo steel tube bundle by intermediate leaks are being prepared

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

Investigation of a steam generator tube rupture sequence using VICTORIA

International Nuclear Information System (INIS)

Bixler, N.E.; Erickson, C.M.; Schaperow, J.H.

1995-01-01

VICTORIA-92 is a mechanistic computer code for analyzing fission product behavior within the reactor coolant system (RCS) during a severe reactor accident. It provides detailed predictions of the release of radionuclides and nonradioactive materials from the core and transport of these materials within the RCS. The modeling accounts for the chemical and aerosol processes that affect radionuclide behavior. Coupling of detailed chemistry and aerosol packages is a unique feature of VICTORIA; it allows exploration of phenomena involving deposition, revaporization, and re-entrainment that cannot be resolved with other codes. The purpose of this work is to determine the attenuation of fission products in the RCS and on the secondary side of the steam generator in an accident initiated by a steam generator tube rupture (SGTR). As a class, bypass sequences have been identified in NUREG-1150 as being risk dominant for the Surry and Sequoyah pressurized water reactor (PWR) plants

Steam generator leak detection using acoustic method

International Nuclear Information System (INIS)

Goluchko, V.V.; Sokolov, B.M.; Bulanov, A.N.

1982-05-01

The main requirements to meet by a device for leak detection in sodium - water steam generators are determined. The potentialities of instrumentation designed based on the developed requirements have been tested using a model of a 550 kw steam generator [fr

Intentional back flow effects on ruptured steam generator cooldown during a SGTR event for KSNP

International Nuclear Information System (INIS)

Kim, C.W.; Park, S.J.; Choi, C.J.; Seo, J.T.

2004-01-01

For an optimum recovery from a steam generator tube rupture (SGTR) event, the operators are directed to isolate the steam generator (SG) with ruptured tube as early as possible to minimize the radioactive material release. However, the reactor coolant system (RCS) cooldown and depressurization to the shutdown cooling system (SCS) operation conditions using the intact SG only are hard to achieve unless the ruptured SG is properly cooled since the ruptured SG, which is isolated by operator, remains at high temperature even though the RCS has been cooled down. The effects of intentional back flow from the SG secondary side to the RCS through the ruptured U-tube on the the ruptured SG cooldown were evaluated for the pressurized light water reactor, especially for the Korean standard nuclear power plant (KSNP). In order to evaluate the back flow effect, a series of analyses was conducted using the RELAP5/MOD3 computer code. For the first stage of the analysis, the cooldown process by natural circulation in the SG secondary side was simulated for the initial conditions of the ruptured SG cooldown. In the next analysis stage, two methods of the ruptured SG cooldown by using back flow after RCS cooldown were evaluated. One utilizes the steam condensation on the uncovered U-tube surface, and the other is a SG drain and fill. In the former method, SG tubes are exposed to the steam space by draining SG secondary water into the RCS in order to condense the steam directly onto the uncovered tubes. This method showed that the steam condensation decreased SG secondary pressure and temperature rapidly, demonstrating its effectiveness for cooling. However, this process has a limited applicability if the rupture is located at the lower region. The latter method, draining by back flow and filling using the feedwater system was also found to be effective in ruptured SG cooldown and depressurization even if the rupture occurred at the top of the U-tube. It is concluded that the

Steam generator asset management: integrating technology and asset management

International Nuclear Information System (INIS)

Shoemaker, P.; Cislo, D.

2006-01-01

, and nuclear fuel, are the most costly components within a nuclear steam supply system, yet they don't lend themselves well to asset management practices. However, application of asset management principles to these components represents the largest potential return for asset management within nuclear generation and sustaining continued plant operation, to include life extensions. Recognizing this critical gap, AREVA launched a Steam Generator Secondary Side Vision to develop advanced technologies for the management of nuclear steam generators. These technologies include advanced mitigation techniques, to include non-invasive techniques, new inspection techniques, predictive modeling and expert systems aimed at minimizing and managing steam generator corrosion and risks. (author)

Mathematical modeling of control system for the experimental steam generator

Directory of Open Access Journals (Sweden)

Podlasek Szymon

2016-01-01

Full Text Available A steam generator is an essential unit of each cogeneration system using steam machines. Currently one of the cheapest ways of the steam generation can be application of old steam generators came from army surplus store. They have relatively simple construction and in case of not so exploited units – quite good general conditions, and functionality of mechanical components. By contrast, electrical components and control systems (mostly based on relay automatics are definitely obsolete. It is not possible to use such units with cooperation of steam bus or with steam engines. In particular, there is no possibility for automatically adjustment of the pressure and the temperature of the generated steam supplying steam engines. Such adjustment is necessary in case of variation of a generator load. The paper is devoted to description of improvement of an exemplary unit together with construction of the measurement-control system based on a PLC. The aim was to enable for communication between the steam generator and controllers of the steam bus and steam engines in order to construction of a complete, fully autonomic and maintenance-free microcogeneration system.

Mechanical design of a sodium heated steam generator

International Nuclear Information System (INIS)

Chetal, S.C.

1975-01-01

FBTR steam generator is a once through type unit consisting of four 12.5 MW thermal modules generating a total of 74 tons per hour of steam at 125 bar and 480 0 C. This paper outlines the mechanical design of such type of steam generator with emphasis on special design problems associated with this type of sodium to water steam heat exchanger, namely, thermal cycling of transition zone where nucleate boiling changes over to film boiling, application of pressure vessel design criteria for transient pressures, thermal stress evaluation resulting from differential expansion between shell and tube in this typical configuration, sodium headers support design, thermal sleeve, design, thermal shock analysis in thick tubes, thermal stress resulting from stratification and stability of expansion bends against vibration. Some of the possible design changes for the future large size steam generator are outlined. (author)

Chemical-Cleaning Demonstration Test No. 2 in a mock-up steam generator

International Nuclear Information System (INIS)

Jevec, J.M.; Leedy, W.S.

1983-04-01

This report describes the results of the mockup demonstration test of the first modified baseline process under Contract S-127, Chemical Cleaning of Nuclear Steam Generators. The objective of this program is to determine the feasibility of cleaning the secondary side of nuclear steam generators with state-of-the-art chemical cleaning technology. The first step was to benchmark a baseline process. This process was then modified to attempt to eliminate the causes of unacceptable cleaning performance. The modified baseline process consists of an EDTA/H 2 O 2 -based copper solvent and a near-neutral, EDTA/N 2 H 4 -based magnetite and crevice solvent. This report also presents the results of three inhibitor evaluation mockup runs used in the evaluation of the modified baseline process

Consolidated nuclear steam generator

International Nuclear Information System (INIS)

Jabsen, F.S.; Schluderberg, D.C.; Paulson, A.E.

1978-01-01

An improved system of providing power has a unique generating means for nuclear reactors with a number of steam generators in the form of replaceable modular units of the expendable type to attain the optimum in effective and efficient vaporization of fluid during the generating power. The system is most adaptable to undrground power plants and marine usage

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

LMFBR steam generators in the United Kingdom

International Nuclear Information System (INIS)

Anderson, R.; Hayden, O.

2002-01-01

Experience has been gained in the UK on the operation of LMFBR Steam Generator Units (SGU) over a period of 20 years from the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR). The DFR steam generator featured a double barrier and therefore did not represent a commercial design. PFR, however, faced the challenge of a single wall design and it is experience from this which is most valuable. The PFR reactor went critical in March 1974 and the plant operating history since then has been dominated by experience with leaks in the tube to tube plate welds of the high performance U-tubes SGU's. Operation at high power using the full complement of three secondary sodium circuits was delayed until July 1976 by the occurrence of leaks in the tube to tube plate welds of the superheater and reheater units which are fabricated in stainless steel. Repairs were carried out to the two superheaters and they were returned to service. The reheater tube bundle was removed from circuit after sodium was found to have entered the steam side. When the sodium had been removed and inspection carried out it was decided not to recover the unit. Since 1976 the remaining five stainless steel units have operated satisfactorily. This year a replacement reheater unit has been installed. This is of a new design in 9-Cr-Mo ferritic steel using a sleeve through which the steam tube passes to eliminate the tube to tube plate weld. Despite a few early leaks in evaporator tube to tube plate welds up to 1979, these failures did not initially present a major problem. However, in 1980 the rate of evaporator weld failures increased and despite the successful application of a shot peening process to eliminate stress corrosion failures from the water side of the weld, failures traced to the sodium side continued. A sleeving process was developed for application to complete evaporator units on a production basis with the objective of bypassing the welds at each end of the 500 tubes. The decision

Steam generator tube integrity program. Phase I report

International Nuclear Information System (INIS)

Alzheimer, J.M.; Clark, R.A.; Morris, C.J.; Vagins, M.

1979-09-01

The results are presented of the pressure tests performed as part of Phase I of the Steam Generator Tube Integrity (SGTI) program at Battelle Pacific Northwest Laboratory. These tests were performed to establish margin-to-failure predictions for mechanically defected Pressurized Water Reactor (PWR) steam generator tubing under operating and accident conditions. Defect geometries tested were selected because they simulate known or expected defects in PWR steam generators. These defect geometries are Electric Discharge Machining (EDM) slots, elliptical wastage, elliptical wastage plus through-wall slot, uniform thinning, denting, denting plus uniform thinning, and denting plus elliptical wastage. All defects were placed in tubing representative of that currently used in PWR steam generators

Double wall steam generator tubing

International Nuclear Information System (INIS)

Padden, T.R.; Uber, C.F.

1983-01-01

Double-walled steam generator tubing for the steam generators of a liquid metal cooled fast breeder reactor prevents sliding between the surfaces due to a mechanical interlock. Forces resulting from differential thermal expansion between the outer tube and the inner tube are insufficient in magnitude to cause shearing of base metal. The interlock is formed by jointly drawing the tubing, with the inside wall of the outer tube being already formed with grooves. The drawing causes the outer wall of the inner tube to form corrugations locking with the grooves. (author)

Experimental verification of the horizontal steam generator boil-off transfer degradation at natural circulation

Energy Technology Data Exchange (ETDEWEB)

Hyvaerinen, J. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

1997-12-31

The presentation summarises the highlights of experimental results obtained for VVER type horizontal steam generator heat transfer, primary side flow pattern, and mixing in the hot collector during secondary side boil-off with primary at single-phase natural circulation. The experiments were performed using the PACTEL facility with Large Diameter (LD) steam generator models, with collector instrumentation designed specifically for these tests. The key findings are as follows: (1) the primary to secondary heat transfer degrades as the secondary water inventory is depleted, following closely the wetted tube area; (2) a circulatory flow pattern exists in the tube bundle, resulting in reversed flow (from cold to the hot collector) in the lower part of the tube bundle, and continuous flow through the upper part, including the tubes that have already dried out; and (3) mixing of the hot leg flow entering the hot collector and reversed, cold, tube flow remains confined within the collector itself, extending only a row or two above the elevation at which tube flow reversal has taken place. 6 refs.

Experimental verification of the horizontal steam generator boil-off transfer degradation at natural circulation

Energy Technology Data Exchange (ETDEWEB)

Hyvaerinen, J [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Kouhia, J [VTT Energy, Lappeenranta (Finland)

1998-12-31

The presentation summarises the highlights of experimental results obtained for VVER type horizontal steam generator heat transfer, primary side flow pattern, and mixing in the hot collector during secondary side boil-off with primary at single-phase natural circulation. The experiments were performed using the PACTEL facility with Large Diameter (LD) steam generator models, with collector instrumentation designed specifically for these tests. The key findings are as follows: (1) the primary to secondary heat transfer degrades as the secondary water inventory is depleted, following closely the wetted tube area; (2) a circulatory flow pattern exists in the tube bundle, resulting in reversed flow (from cold to the hot collector) in the lower part of the tube bundle, and continuous flow through the upper part, including the tubes that have already dried out; and (3) mixing of the hot leg flow entering the hot collector and reversed, cold, tube flow remains confined within the collector itself, extending only a row or two above the elevation at which tube flow reversal has taken place. 6 refs.

Experimental verification of the horizontal steam generator boil-off transfer degradation at natural circulation

International Nuclear Information System (INIS)

Hyvaerinen, J.; Kouhia, J.

1997-01-01

The presentation summarises the highlights of experimental results obtained for VVER type horizontal steam generator heat transfer, primary side flow pattern, and mixing in the hot collector during secondary side boil-off with primary at single-phase natural circulation. The experiments were performed using the PACTEL facility with Large Diameter (LD) steam generator models, with collector instrumentation designed specifically for these tests. The key findings are as follows: (1) the primary to secondary heat transfer degrades as the secondary water inventory is depleted, following closely the wetted tube area; (2) a circulatory flow pattern exists in the tube bundle, resulting in reversed flow (from cold to the hot collector) in the lower part of the tube bundle, and continuous flow through the upper part, including the tubes that have already dried out; and (3) mixing of the hot leg flow entering the hot collector and reversed, cold, tube flow remains confined within the collector itself, extending only a row or two above the elevation at which tube flow reversal has taken place

Stress analysis of HTR-10 steam generator heat exchanging tubes

International Nuclear Information System (INIS)

Dong Jianling; Zhang Xiaohang; Yin Dejian; Fu Jiyang

2001-01-01

Steam Generator (SG) heat exchanging tubes of 10 MW High Temperature Gas Cooled Reactor (HTR-10) are protective screens between the primary loop of helium with radioactivity and the secondary loop of feeding water and steam without radioactivity. Water and steam will enter into the primary loop when rupture of the heat exchanging tubes occurs, which lead to increase of the primary loop pressure and discharge of radioactive materials. Therefore it is important to guarantee the integrity of the tubes. The tube structure is spiral tube with small bending radius, which make it impossible to test with volumetric in-service detection. For such kind of spiral tube, using LBB concept to guarantee the integrity of the tubes is an important option. The author conducts stress analysis and calculation of HTR-10 SG heat exchanging tubes using the FEM code of piping stress analysis, PIPESTRESS. The maximum stress and the dangerous positions are obtained

Coal fired steam generation for heavy oil recovery

International Nuclear Information System (INIS)

Firmin, K.

1992-01-01

In Alberta, some 21,000 m 3 /d of heavy oil and bitumen are produced by in-situ recovery methods involving steam injection. The steam generation requirement is met by standardized natural-gas-fired steam generators. While gas is in plentiful supply in Alberta and therefore competitively priced, significant gas price increases could occur in the future. A 1985 study investigating the alternatives to natural gas as a fuel for steam generation concluded that coal was the most economic alternative, as reserves of subbituminous coal are not only abundant in Alberta but also located relatively close to heavy oil and bitumen production areas. The environmental performance of coal is critical to its acceptance as an alternate fuel to natural gas, and proposed steam generator designs which could burn Alberta coal and control emissions satisfactorily are assessed. Considerations for ash removal, sulfur dioxide sorption, nitrogen oxides control, and particulate emission capture are also presented. A multi-stage slagging type of coal-fired combustor has been developed which is suitable for application with oilfield steam generators and is being commissioned for a demonstration project at the Cold Lake deposit. An economic study showed that the use of coal for steam generation in heavy oil in-situ projects in the Peace River and Cold Lake areas would be economic, compared to natural gas, at fuel price projections and design/cost premises for a project timing in the mid-1990s. 7 figs., 3 tabs

Stress analysis of steam generator row-1 tubes

International Nuclear Information System (INIS)

Kim, Woo Gon; Ryu, Woo Seog; Lee, Ho Jin; Kim, Sung Chung

2000-01-01

Residual stresses induced in U-bending and tube-to-tubesheet joining processes of PWR's steam generator row-1 tube were measured by X-ray method and Hole-Drilling Method(HDM). The stresses resulting from the internal pressure and the temperature gradient in the steam generator were also estimated theoretically. In U-bent regions, the residual stresses at extrados were induced with compressive stress(-), and its maximum value reached -319 Mpa in axial direction at ψ=0 .deg. in position. Maximum tensile residual stress of 170 MPa was found to be at the flank side at position of ψ=90 deg., i.e., at apex region. In tube-to-tubesheet joining methods, the residual stresses induced by the explosive joint method were found to be lower than that by the mechanical roll method. The gradient of residual stress along the expanded tube was highest at the transition region, and the residual stress in circumferential direction was found to be higher than the residual stress in axial direction. Hoop stress due to an internal pressure between primary and secondary side was analyzed to be 76 MPa and thermal stress was 45 MPa

Forming a cohesive steam generator maintenance strategy

International Nuclear Information System (INIS)

Poudroux, G.

1991-01-01

In older nuclear plants, steam generator tube bundles are the most fragile part of the reactor coolant system. Steam generator tubes are subject to numerous types of loading, which can lead to severe degradation (corrosion and wear phenomena). Preventive actions, such as reactor coolant temperature reduction or increasing the plugging limit and their associated analyses, can increase steam generator service life. Beyond these preventive actions, the number of affected tubes and the different locations of the degradations that occur often make repair campaigns necessary. Framatome has developed and qualified a wide range of treatment and repair processes. They enable careful management of the repair campaigns, to avoid reaching the maximum steam generator tube plugging limit, while optimizing the costs. Most of the available repair techniques allow a large number of affected tubes to be treated. Here we look only at those techniques that should be taken into account when defining a maintenance strategy. (author)

Probabilistic methodology for assessing steam generator tube inspection - Phase II: User's manual for CANTIA Version 1.1

International Nuclear Information System (INIS)

Harris, J.E.; Gorman, J.A.; Turner, A.P.L.

1997-03-01

The objectives of this project were to develop a computer-based method for probabilistic assessment of inspection strategies for steam generator tubes, and to document the source code and to provide a user's manual for it. The program CANTIA was created to fulfill this objective, and the user's manual is provided in this volume. The documentation and verification of the CANTIA code is provided as a separate report. CANTIA uses Monte Carlo techniques to determine approximate probabilities of steam generator tube failures under accident conditions and primary-to-secondary leak rates under normal and accident conditions at future points in time. The program also determines approximate future flaw distributions and non-destructive examination results from the input data. The probabilities of failure and leak rates and the future flaw distributions can be influenced by performing inspections of the steam generator tubes at some future points in time, and removing defective tubes from the population. The effect of different inspection and maintenance strategies can therefore be determined as a direct effect on the probability of tube failure and primary-to-secondary leak rate

Susceptibility of CANDU steam generator preheater to cavitation erosion

International Nuclear Information System (INIS)

Laroche, S.L.; Sun, L.; Pietralik, J.M.

2012-01-01

In 2009, Darlington Steam Generator (SG) tube inspections revealed some tubes had degraded in the preheater. The tube degradation occurred at the clearance gap between the tube and the preheater baffle and reached up to 50% through-wall depth at the baffles in the middle portion of the preheater. The general pattern of the damage and the elemental composition analysis suggested that the degradation was the result of a hydrodynamic process, such as cavitation erosion. Cavitation erosion occurs when vapour bubbles exist or form in the flowing liquid and then these bubbles collapse violently in the vicinity of the wall. These bubbles collapse when steam bubbles contact water that is sufficiently subcooled, below the saturation temperature. In the gap between the tube and the preheater baffle, low flow will exist due to the pressure difference across the baffle plate. In addition, heat transfer occurs from the primary-side fluid to the secondary-side fluid within this clearance gap that is driven by the primary-to-secondary temperature difference. Factors, such as the tube position in the baffle hole and fouling, influence the local conditions and can cause subcooled boiling that result in cavitation. This paper presents a study of flow and heat transfer phenomena to determine the factors contributing to cavitation erosion in SG preheaters. The analysis used the THIRST1 code for a 3-dimensional thermalhydraulic simulation of the steam generators and the ANSYS FLUENT®2 code for detailed calculations of flow and heat transfer in the clearance gaps. This study identifies that tubes in the preheater region are susceptible to cavitation erosion and indicates that this area should be part of the station inspection program because, regardless of preheater design, some tubes may experience the thermalhydraulic conditions and undergo degradations similar to those observed for the tubes in Darlington SGs. (author)

Condensate polisher application for PWR steam generator corrosion control

International Nuclear Information System (INIS)

Sawochka, S.G.; Leibovitz, J.; Siegwarth, D.P.; Pearl, W.L.

1981-01-01

The evolution of corrosion attack modes particularly in recirculating U-tube PWR steam generators has dictated a thorough review of the advantages and disadvantages of condensate polishing. Analytical modeling techniques to qualitatively predict crevice chemistry variations resulting from steam generator bulk water variations have allowed valuable insights to be developed. Modeling results complemented by steam generator and laboratory corrosion data will be employed to set condensate demineralizer effluent specifications consistent with control of steam generator corrosion. Laboratory and plant studies are being performed to demonstrate achievability of necessary effluent specifications. (author)

Babcock and Wilcox Canada steam generators past, present and future

Energy Technology Data Exchange (ETDEWEB)

Smith, J.C. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada)

1998-07-01

The steam generators in all of the domestic CANDU Plants, and most of the foreign CANDU plants, were supplied by Babcock and Wilcox Canada, either on their own or in co-operation with local manufacturers. More than 200 steam generators have been supplied. In addition, Babcock and Wilcox Canada has taken the technology which evolved out of the CANDU steam generators and has adapted the technology to supply of replacement steam generators for PWR's. There is enough history and operating experience, plus laboratory experience, to point to the future directions which will be taken in steam generator design. This paper documents the steam generators which have been supplied, the experience in operation and maintenance, what has worked and not worked, and how the design, materials, and operating and maintenance philosophy have evolved. The paper also looks at future requirements in the market, and the continuing research and product development going on at Babcock and Wilcox to address the future steam generator requirements. (author)

Babcock and Wilcox Canada steam generators past, present and future

International Nuclear Information System (INIS)

Smith, J.C.

1998-01-01

The steam generators in all of the domestic CANDU Plants, and most of the foreign CANDU plants, were supplied by Babcock and Wilcox Canada, either on their own or in co-operation with local manufacturers. More than 200 steam generators have been supplied. In addition, Babcock and Wilcox Canada has taken the technology which evolved out of the CANDU steam generators and has adapted the technology to supply of replacement steam generators for PWR's. There is enough history and operating experience, plus laboratory experience, to point to the future directions which will be taken in steam generator design. This paper documents the steam generators which have been supplied, the experience in operation and maintenance, what has worked and not worked, and how the design, materials, and operating and maintenance philosophy have evolved. The paper also looks at future requirements in the market, and the continuing research and product development going on at Babcock and Wilcox to address the future steam generator requirements. (author)

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)

Chemical cleaning - essential for optimal steam generator asset management

International Nuclear Information System (INIS)

Ammann, Franz

2009-01-01

Accumulation of deposits in Steam Generator is intrinsic during the operation of Pressurized Water Reactors. Such depositions lead to reduction of thermal performance, loss of component integrity and, in some cases, to power restrictions. Accordingly, removal of such deposits is an essential part of the asset management program of Steam Generators. Every plant has specific conditions, history and constraints which must be considered when planning and performing a chemical cleaning. Typical points are: -Constitution of the deposits or sludge - Sludge load - Sludge distribution in the steam generator - Existing or expected corrosion problems - Amount and tendency of fouling for waste treatment The strategy for chemical cleaning is developed from these points. The range of chemical cleaning treatments starts with very soft cleanings which can remove approximately 100kg per steam generator and ends with full scale, i.e., hard, cleanings which can remove several thousand kilograms of deposits from a steam generator. Dependent upon the desired goal for the operating plant and the steam generator material condition, the correct cleaning method can be selected. This requires flexible cleaning methods that can be adapted to the individual needs of a plant. Such customizing of chemical cleaning methods is a crucial factor for an optimized asset management program of steam generators in a nuclear power plant

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

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

Optimum fuel allocation in parallel steam generator systems

International Nuclear Information System (INIS)

Bollettini, U.; Cangioli, E.; Cerri, G.; Rome Univ. 'La Sapienza'; Trento Univ.

1991-01-01

An optimization procedure was developed to allocate fuels into parallel steam generators. The procedure takes into account the level of performance deterioration connected with the loading history (fossil fuel allocation and maintenance) of each steam generator. The optimization objective function is the system hourly cost, overall steam demand being satisfied. Costs are due to fuel and electric power supply and to plant depreciation and maintenance as well. In order to easily updata the state of each steam generator, particular care was put in the general formulation of the steam production function by adopting a special efficiency-load curve description based on a deterioration scaling parameter. The influence of the characteristic time interval length on the optimum operation result is investigated. A special implementation of the method based on minimum cost paths is suggested

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.

Two-phase flow field simulation of horizontal steam generators

Energy Technology Data Exchange (ETDEWEB)

Rabiee, Ataollah; Kamalinia, Amir Hossein; Hadad, Kamal [School of Mechanical Engineering, Shiraz University, Shiraz (Iran, Islamic Republic of)

2017-02-15

The analysis of steam generators as an interface between primary and secondary circuits in light water nuclear power plants is crucial in terms of safety and design issues. VVER-1000 nuclear power plants use horizontal steam generators which demand a detailed thermal hydraulics investigation in order to predict their behavior during normal and transient operational conditions. Two phase flow field simulation on adjacent tube bundles is important in obtaining logical numerical results. However, the complexity of the tube bundles, due to geometry and arrangement, makes it complicated. Employment of porous media is suggested to simplify numerical modeling. This study presents the use of porous media to simulate the tube bundles within a general-purpose computational fluid dynamics code. Solved governing equations are generalized phase continuity, momentum, and energy equations. Boundary conditions, as one of the main challenges in this numerical analysis, are optimized. The model has been verified and tuned by simple two-dimensional geometry. It is shown that the obtained vapor volume fraction near the cold and hot collectors predict the experimental results more accurately than in previous studies.

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

Research perspectives on the evaluation of steam generator tube integrity

International Nuclear Information System (INIS)

Muscara, J.; Diercks, D. R.; Majumdar, S.; Kupperman, D. S.; Bakhtiari, S.; Shack, W. J.

2001-01-01

Industry effects have been largely successful in managing degradation of steam generator tubes due to wastage, pitting, and denting, but fretting, SCC and intergranular attack have proved more difficult to manage. Although steam generator replacements are proceeding there is substantial industry interest in operating with degraded steam generators, and significant numbers of plants will continue to do so. In most cases degradation of steam generator tubing by stress corrosion cracking is still managed by plug or repair on detection, because current NDE techniques for characterization of flaws are not accurate enough to permit continued operation. This paper reviews some of the historical background that underlies current steam generator degradation management strategies and outlines some of the additional research that must be done to provide more effective management of degradation in current generators and provide greater assurance of satisfactory performance in replacement steam generators

SNR-300 steam generator accident philosophy - Assessment due to new understandings in Na/H20-reactions

International Nuclear Information System (INIS)

Ruloff, G.; Huebner, R.

1990-01-01

Recent R+D results in the intermediate leak range (finally confirmed by the PFR steam generator accident) lead to a new assessment for the SNR-300 steam generator accident. This paper discusses the course of such accident which has to be expected under the SNR-300 conditions, starting with an unblocked micro leak and ending with the pressure loads on the secondary system due to overheating failure. Also, enclosed are the possibilities for a leak detection before serious damage has occurred and the discussion of the definition of the DBA. (author). 2 refs, 9 figs

Highly Flexible and Efficient Solar Steam Generation Device.

Science.gov (United States)

Chen, Chaoji; Li, Yiju; Song, Jianwei; Yang, Zhi; Kuang, Yudi; Hitz, Emily; Jia, Chao; Gong, Amy; Jiang, Feng; Zhu, J Y; Yang, Bao; Xie, Jia; Hu, Liangbing

2017-08-01

Solar steam generation with subsequent steam recondensation has been regarded as one of the most promising techniques to utilize the abundant solar energy and sea water or other unpurified water through water purification, desalination, and distillation. Although tremendous efforts have been dedicated to developing high-efficiency solar steam generation devices, challenges remain in terms of the relatively low efficiency, complicated fabrications, high cost, and inability to scale up. Here, inspired by the water transpiration behavior of trees, the use of carbon nanotube (CNT)-modified flexible wood membrane (F-Wood/CNTs) is demonstrated as a flexible, portable, recyclable, and efficient solar steam generation device for low-cost and scalable solar steam generation applications. Benefitting from the unique structural merits of the F-Wood/CNTs membrane-a black CNT-coated hair-like surface with excellent light absorbability, wood matrix with low thermal conductivity, hierarchical micro- and nanochannels for water pumping and escaping, solar steam generation device based on the F-Wood/CNTs membrane demonstrates a high efficiency of 81% at 10 kW cm -2 , representing one of the highest values ever-reported. The nature-inspired design concept in this study is straightforward and easily scalable, representing one of the most promising solutions for renewable and portable solar energy generation and other related phase-change applications. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.