Operating performance of CANDU pressure tubes

International Nuclear Information System (INIS)

Cheadle, B.A.; Price, E.G.

1989-04-01

The performance of Zircaloy-2 and Zr-2.5 Nb pressure tubes in CANDU reactors is reviewed. The accelerated hydriding of Zircaloy-2 in reducing water chemistries can lower the toughness of this material and it is essential that defect-initiating phenomena, such as hydride blister formation from pressure tube to calandria tube contact, be prevented. Zr-2.5 Nb pressure tubes are performing well with low rates of hydrogen pick-up and good retention of material properties

Deuterium absorption in CANDU Zr-2.5Nb pressure tubes

International Nuclear Information System (INIS)

Ploc, R.A.; McRae, G.A.

1999-12-01

Corrosion of CANDU Zr-2.5%Nb pressure tubes in heavy water results in the formation of an oxide film and the absorption of deuterium by the alloy. If deuterium concentrations are allowed to exceed the terminal solid solubility of the alloy, brittle deuterides can form, thereby limiting the service life of a component. In CANDU pressure tubes, ingress rates are largely determined by the metastable β-Zr that is present as a thin layer encasing the predominant α-Zr grains (approximately 90% by volume). The distribution and continuity of the corroded β-phase in the oxide provides a pervasive web for the development of interconnected porosity from the free surface to the oxide/metal interface. Changing the distribution of the β-phase in the alloy changes the nature of the oxide porosity, a technique that can be used to reduce deuterium ingress rates. (author)

Deuterium absorption in CANDU Zr-2.5Nb pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Ploc, R.A.; McRae, G.A

1999-12-01

Corrosion of CANDU Zr-2.5%Nb pressure tubes in heavy water results in the formation of an oxide film and the absorption of deuterium by the alloy. If deuterium concentrations are allowed to exceed the terminal solid solubility of the alloy, brittle deuterides can form, thereby limiting the service life of a component. In CANDU pressure tubes, ingress rates are largely determined by the metastable {beta}-Zr that is present as a thin layer encasing the predominant {alpha}-Zr grains (approximately 90% by volume). The distribution and continuity of the corroded {beta}-phase in the oxide provides a pervasive web for the development of interconnected porosity from the free surface to the oxide/metal interface. Changing the distribution of the {beta}-phase in the alloy changes the nature of the oxide porosity, a technique that can be used to reduce deuterium ingress rates. (author)

Development of delayed hydride cracking resistant-pressure tube

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Kwon, Sang Chul; Kim, S. S.; Yim, K. S

2000-10-01

For the first time, we demonstrate that the pattern of nucleation and growth of a DHC crack is governed by the precipitation of hydrides so that the DHC velocity and K{sub IH} are determined by an angle of the cracking plane and the hydride habit plane 10.7. Since texture controls the distribution of the 10.7 habit plane in Zr-2.5Nb pressure tube, we draw a conclusion that a textural change in Zr-2.5Nb tube from a strong tangential texture to the radial texture shall increase the threshold stress intensity factor, K{sub IH}, and decrease the delayed hydride cracking velocity. This conclusion is also verified by a complimentary experiment showing a linear dependence of DHCV and K{sub IH} with an increase in the basal component in the cracking plane. On the basis of the study on the DHC mechanism and the effect of manufacturing processes on the properties of Zr-2.5Nb tube, we have established a manufacturing procedure to make pressure tubes with improved DHC resistance. The main features of the established manufacturing process consist in the two step-cold pilgering process and the intermediate heat treatment in the {alpha} + {beta} phase for Zr-2.5Nb alloy and in the {alpha} phase for Zr-1Nb-1.2Sn-0.4Fe alloy. The manufacturing of DHC resistant-pressure tubes of Zr-2.5Nb and Zr-1N-1.2Sn-0.4Fe was made in the ChMP zirconium plant in Russia under a joint research with Drs. Nikulina and Markelov in VNIINM (Russia). Zr-2.5Nb pressure tube made with the established manufacturing process has met all the specification requirements put by KAERI. Chracterization tests have been jointly conducted by VNIINM and KAERI. As expected, the Zr-2.5Nb tube made with the established procedure has improved DHC resistance compared to that of CANDU Zr-2.5Nb pressure tube used currently. The measured DHC velocity of the Zr-2.5Nb tube meets the target value (DHCV <5x10{sup -8} m/s) and its other properties also were equivalent to those of the CANDU Zr-2.5Nb tube used currently. The Zr-1Nb-1

Slit-burst testing of cold-worked Zr-2.5 wt.% Nb pressure tubing for CANDU-PHW reactors

International Nuclear Information System (INIS)

Wilkins, B.J.S.; Barrie, J.N.; Zink, R.J.

1978-12-01

This report documents the available data on critical crack length of cold-worked Zr-2.5 wt.% Nb pressure tubing in CANDU reactors. In particular, it includes data for tubing removed from the Pickering 3 and 4 reactors. (author)

Life management of Zr 2.5% Nb pressure tube through estimation of fracture properties by cyclic ball indentation technique

International Nuclear Information System (INIS)

Chatterjee, S.; Madhusoodanan, K.; Rama Rao, A.

2015-01-01

In Pressurised Heavy Water Reactors (PHWRs) fuel bundles are located inside horizontal pressure tubes. Pressure tubes made up of Zr 2.5 wt% Nb undergo degradation during in-service environmental conditions. Measurement of mechanical properties of degraded pressure tubes is important for assessing its fitness for further service in the reactor. The only way to accomplish this important objective is to develop a system based on insitu measurement technique. Considering the importance of such measurement, an In-situ Property Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed indigenously. The remotely operable system is capable of carrying out indentation trial on the inside surface of the pressure tube and to estimate important mechanical properties like yield strength, ultimate tensile strength, hardness etc. It is known that fracture toughness is one of the important life limiting parameters of the pressure tube. Hence, five spool pieces of Zr 2.5 wt% Nb pressure tube of different mechanical properties have been used for estimation of fracture toughness by ball indentation method. Curved Compact Tension (CCT) specimens were also prepared from the five spool pieces for measurement of fracture toughness from conventional tests. The conventional fracture toughness values were used as reference data. A methodology has been developed to estimate the fracture properties of Zr 2.5 wt% Nb pressure tube material from the analysis of the ball indentation test data. This paper highlights the comparison between tensile properties measured from conventional tests and IProMS trials and relates the fracture toughness parameters measured from conventional tests with the IProMS estimated fracture properties like Indentation Energy to Fracture. (author)

Delayed hydride cracking in Zr-2.5% wt Nb pressure tubes

International Nuclear Information System (INIS)

Cirimello, Pablo; Haddad, Roberto; Domizzi, Gladys

2003-01-01

During service, pressure tubes of CANDU nuclear power reactor are prone to suffer crack growth by delayed hydride cracking (DHC). For a given H 2 plus D 2 concentration there is a critical temperature (T c ) below which DHC may occur. In this work, T c was measured for CCT specimens cut from Zr-2.5 Wt % Nb pressure tubes. Hydrogen was added to the specimens to get concentrations of 40, 59 and 72 ppm. It was found that T c is higher than the corresponding precipitation temperature. The axial crack velocity (V p ) was also measured. Decreasing temperature from T c makes V p increase until a maximum is attained at a temperature close to precipitation temperature. At lower temperatures, in the presence of precipitated hydrides, decreasing temperature implies lower velocities, following an Arrhenius law: Vp=Aexp(-Q/RT), with an activation energy Q= 66 KJ/mol K. (author)

Statistical analysis and modelling of in-reactor diametral creep of Zr-2.5Nb pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Jyrkama, Mikko I., E-mail: mjyrkama@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada); Bickel, Grant A., E-mail: grant.bickel@cnl.ca [Canadian Nuclear Laboratories, Chalk River Laboratories, Chalk River, ON, Canada K0J 1J0 (Canada); Pandey, Mahesh D., E-mail: mdpandey@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada)

2016-04-15

Highlights: • New and simple statistical model of pressure tube diametral creep. • Based on surveillance data of 328 pressure tubes from eight different CANDU reactors. • Uses weighted least squares (WLS) to regress out operating conditions. • The shape of the diametral creep profiles are predicted very well. • Provides insight and relative ranking of strain behaviour of in-service tubes. - Abstract: This paper presents the development of a simplified regression approach for modelling the diametral creep over time in Zr-2.5 wt% Nb pressure tubes used in CANDU reactors. The model is based on a large dataset of in-service inspection data of 328 different pressure tubes from eight different CANDU reactor units. The proposed weighted least squares (WLS) regression model is linear in time as a function of flux and temperature, with a temperature-dependent variance function. The model predicts the shape of the observed diametral creep profiles very well, and is useful not merely for prediction, but also for assessing tube-to-tube variability and manufacturing properties among the inspected tubes.

Fatigue crack initiation at complex flaws in hydrided Zr-2.5%Nb samples from CANDU pressure tubes

International Nuclear Information System (INIS)

Stoica, L.; Radu, V.

2016-01-01

The paper addresses the phenomena which occur at locations where the oxide layer of the inner surface of CANDU tube pressure is damaged by the contact with the fuel element or due to the action of hard particles at the interface between the tube pressure and bearing pad of fuel element. In such situations generate defects, which most often are defects known as ''bearing pad fretting flaws'' or ''debris fretting flaws''. In this paper the experiments are completed in a series of previous works on the mechanical fatigue phenomenon on samples prepared from the pressure tube Zr-2.5% Nb alloy. The phenomenon of variable mechanical stress (or fatigue) may lead to initiation of cracks at the tip of volumetric flaws, according to the accumulation of hydrides, which then fractures and can propagate through the tube wall pressure due to the mechanism of type DHC (Delayed Hydride Cracking). (authors)

Fracture toughness behaviour using small CCT specimen of Zr-2.5Nb pressure tube materials

International Nuclear Information System (INIS)

Oh, Dong Joon; Kim, Young Suk; Ahn, Sang Bok; Im, Kyung Soo; Kwon, Sang Chul; Cheong, Yong Mu

2001-03-01

Fracture toughness of Zr-2.5Nb pressure tube is the essential data to estimate the CCL(critical crack length) for the concept of LBB(Leak-Before-Break) in PHWR. Zr-2.5Nb pressure tubes could be degraded due to the absorption of hydrogen from coolant and the irradiation. To investigate the fracture toughness behaviour such as J-resistance curves, dJ/da, and CCL of some Zr-alloys (CANDU-double, -quad, CW-E125, TMT-E125, E-635), the transverse tensile test and the fracture toughness test of small CCT (Curved Compact Tension) specimen with 17 mm width were carried out with the variation of testing temperature at different testing condition. To define the fracture mechanism of degradation, the fractographic comparison of fracture surface was performed using the stereoscope and SEM. In addition, the effect of non-uniformed pre-fatigue crack was also studied. In conclusion, CANDU double-melted was less tougher than CANDU quad-melted and the hydrogen embrittlement was found at room temperature. Finally, while the effect of non-uniformed pre-fatigue crack was considerable at room temperature, this effect was disappeared at 250-300 .deg. C

Development of modified route for fabrication of Zr-2.5Nb alloy pressure tubes

International Nuclear Information System (INIS)

Saibaba, N.; Hemantha Rao, G.V.S.; Phani Babu, C.; Jha, S.K.; Ganesha, G.N.; Ramana Rao, S.V.; Kumar Vaibhaw; Dey, G.K.; Srivastava, D.; Neogy, S.; Mani Krishna, K.V.

2013-01-01

Different fabrication trials involving the variation in three important manufacturing stages of Zr-2.5%Nb pressure tube were partially undetaken. The variations were with respect of mode of breaking the cast structure of the ingot (forging vs extrution), ratio of hot extrusion and number of stages of subsequent cold work to produce the finished tube. It was observed that forging process resulted in superior performance in breaking the cast structure. Higher extrusion ratios resulted in more favorable texture and microstrucutre. Continuity of the beta phase in the final microstructure was observed to be more in case of route involving single cold work subsequent to hot extrusion. (author)

Microstructure control of Zr-Nb-Sn alloy with Mo addition for HWR pressure tube application

International Nuclear Information System (INIS)

Hwang, S. K.; Kim, M. H.; Kim, J. H.; Kwon, S. I.; Kim, Y. S.

1997-01-01

As a basic research to develop the material for heavy water reactor pressure tube application the effect of Mo addition to Zr-Nb-Sn alloy was studied for the purpose of minimizing the amount of cold working while maintaining a high strength. To select the target alloy system we first designed various alloy compositions and chose Zr-Nb-Sn and Zr-Nb-Mo through multi-regression analysis of the relationship between the basic properties and the compositions. Plasma arc melting was used to produce the alloys and the microstructure change introduced by the processing steps including hot forging, beta-heat treatment, hot rolling, cold rolling and recrystallization heat treatment was investigated. Recrystallization of Zr-Nb-Sn was retarded by adding Mo and this resulted in a fine grain structure in Zr-Nb-Sn-Mo alloy. Beside the retarding effect recrystallization, Mo increased the amount of residual beta phase and showed an indication of precipitation hardening, which added up to the possibility of applying the alloy for the desired usage. (author)

Influence of hydrogen content on impact toughness of Zr-2.5Nb pressure tube alloy

Energy Technology Data Exchange (ETDEWEB)

Singh, R.N., E-mail: rnsingh@barc.gov.in [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Viswanathan, U.K.; Kumar, Sunil; Satheesh, P.M.; Anantharaman, S. [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Stahle, P. [Division of Solid Mechanics, Lund University/LTH, SE22100 Lund (Sweden)

2011-07-15

Highlights: > For the first time impact behaviour of Zr-2.5Nb pressure tube material used in Indian Pressurized Heavy Water Reactor (IPHWR) as a function of hydrogen content and temperature is being reported. > The critical hydrogen concentration to cause low energy fracture at 25 and 200 deg. C is suggested. > The impact behaviour is rationalized in terms of hydrogen content, test temperature, microstructural features and state of stress ahead of a crack. - Abstract: Influence of hydrogen content on the impact toughness of Zr-2.5% Nb alloy was examined by carrying out instrumented drop weight tests in the temperature range of 25-250 deg. C using curved Charpy specimens fabricated from unirradiated pressure tubes of Indian Pressurized Heavy Water Reactor (IPHWR). Hydrogen content of the samples was between 10 and 170 ppm by weight (wppm). Sharp ductile-to-brittle-transition behaviour was demonstrated by hydrided materials. The temperature for the onset of transition increased with the increase in the hydrogen content of the specimens. The fracture surfaces of unhydrided specimen exhibited ductile fracture caused by micro void coalescence and tear ridges at lower temperatures and by fibrous fracture at intermediate and at higher temperatures. Except for the samples tested at the upper shelf energy levels, the fracture surfaces of all hydrided samples were suggestive of hydride assisted failure. In most cases the transverse cracks observed in the fracture path matched well with the hydride precipitate distribution and orientation.

Performance of pressure tubes in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

DHC velocity comparison of CANDU Zr-2.5Nb pressure tube for the difference of test environments

International Nuclear Information System (INIS)

Cho, S. Y.; Kim, Y. S.; Oh, D. Z.; Yim, K. S.; Yim, Y. W.

2001-01-01

Zr-2.5Nb Pressure tube was used in the distilled water under high temperature and pressure. However, the evaluation of DHCV for pressure tube was limited in the air until now. Therefore, it was necessary for DHCV both in the air and in the distilled water under high temperature and pressure to evaluate. In advance, new DHC equipment simulating the real operating condition in the distilled water under high temperature and pressure was developed and DHCV test was conducted by this equipment. The test was carried out under simulated condition using distilled water of 250 .deg. C, 86bar and this result was compared with of DHCV in the air of 250 .deg. C. DHCV of the distilled water was ranged from 8.42x10 -8 to 9.92x 10 -8 m/s and the average value was 9.01x 10 -8 m/s. As compared with the air condition, it was found that characteristics of DHCV was not affected by the distilled water of high temperature and pressure. At the same temperature, DHCV of the irradiated pressure tube was faster than that of test result

Evaluation of the crack initiation of curved compact tension specimens of a Zr-2.5Nb pressure tube using the unloading compliance and direct current potential drop methods

International Nuclear Information System (INIS)

Jeong, Hyeon Cheol; Ahn, Sang Bok; Park, Joong Chul; Kim, Young Suk

2005-01-01

Zr-2.5Nb pressure tubes, carrying fuel bundles and heavy water coolant inside, degrade due to neutron irradiation and hydrogen embrittlement during their operation in heavy water reactors. The safety criterion for the Zr-2.5Nb tubes to meet is a leak-before-break (LBB) requirement. To evaluate a safety margin related to the LBB criterion, facture toughness of the pressure tubes are to be determined periodically with their operational time. For a reliable evaluation of the LBB safety criterion of the pressure tubes, it is required to precisely determine their fracture toughness. Since the fracture toughness or J of the pressure tubes is determined only by the extended crack length, it is important to reliably and precisely evaluate the advanced crack length. However, the problem lies with the detection of the crack opening point because prior plastic deformation before a start of the crack makes it difficult. The aim of this work is to evaluate which method can define the crack initiation point in the Zr- 2.5Nb compact tension specimens more precisely between the unloading compliance method with a crack opening displacement (COD) gauge and the direct current potential drop (DCPD) methods

Fracture toughness of irradiated Zr-2.5Nb pressure tube from KAPS-2 evaluated using disk compact tension specimens

International Nuclear Information System (INIS)

Shah, Priti Kotak; Dubey, J.S.; Balakrishnan, K.S.; Shriwastaw, R.S.; Dhotre, M.P.; Bhandekar, A.; Pandit, K.M.; Anantharaman, S.

2013-12-01

The report gives the results of the fracture toughness tests carried out over the range of temperatures on specimens prepared from the irradiated S-07Zr-2.5Nb pressure tube removed from Kakrapar Atomic Power Station-2 (KAPS-2) as a part of materials surveillance programme. The pressure tube had experienced ∼ 8 effective full power years (EFPY) of reactor operation and had hydrogen equivalent (H eq ) content less than 20 ppm along the tube length. The fracture toughness tests have been carried out using 30 mm Disk Compact Tension (DCT) specimens, that were punched out of the irradiated pressure tube. The disk punching was carried out using specially made shielded enclosure and hydraulic press. Fatigue pre-cracking and fracture toughness tests were performed using servo-hydraulic universal testing machine with Direct Current Potential Drop (DCPD) equipment to monitor the crack length. The tests were carried out at different test temperature from ambient to 300℃. The fracture toughness values have been used to estimate the critical pressure for the tube. The fracture properties indicate that such tubes have sufficient toughness to satisfy the Leak-Before-Break (LBB) criterion for in-reactor operation. (author)

Measurement of mechanical properties of a reactor operated Zr–2.5Nb pressure tube using an in situ cyclic ball indentation system

Energy Technology Data Exchange (ETDEWEB)

Chatterjee, S., E-mail: subrata@barc.gov.in; Panwar, Sanjay; Madhusoodanan, K.

2015-07-15

Highlights: • Measurement of mechanical properties of pressure tube is required for its fitness assessment. • Pressure tube removal from the core consumes large amount of radiation for laboratory test. • A remotely operable In situProperty Measurement System has been designed in house. • The tool head is capable to carry out in situ ball indentation trials inside pressure tube. • The paper describes the theory and results of the trials conducted on irradiated pressure tube. - Abstract: Periodic measurement of mechanical properties of pressure tubes of Indian Pressurised Heavy Water Reactors is required for assessment of their fitness for continued operation. Removal of pressure tube from the core for preparation of specimens to test for mechanical properties in laboratories consumes large amounts of radiation and hence is to be avoided as far as possible. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing indentation test either on outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ condition. Considering this, a remotely operable hydraulic In situProperty Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed in house. The tool head of IProMS can be located inside a pressure tube at any axial location under in situ condition and the properties can be estimated from an analysis of the data on load and depth of indentation, recorded during the test. In order to qualify the system, a number of experimental trials have been conducted on spool pieces and specimens prepared from Zr–2.5Nb pressure tube having different mechanical properties. Based on the encouraging results obtained from the qualification trials, IProMS has been used inside a reactor operated

Measurement of mechanical properties of a reactor operated Zr–2.5Nb pressure tube using an in situ cyclic ball indentation system

International Nuclear Information System (INIS)

Chatterjee, S.; Panwar, Sanjay; Madhusoodanan, K.

2015-01-01

Highlights: • Measurement of mechanical properties of pressure tube is required for its fitness assessment. • Pressure tube removal from the core consumes large amount of radiation for laboratory test. • A remotely operable In situProperty Measurement System has been designed in house. • The tool head is capable to carry out in situ ball indentation trials inside pressure tube. • The paper describes the theory and results of the trials conducted on irradiated pressure tube. - Abstract: Periodic measurement of mechanical properties of pressure tubes of Indian Pressurised Heavy Water Reactors is required for assessment of their fitness for continued operation. Removal of pressure tube from the core for preparation of specimens to test for mechanical properties in laboratories consumes large amounts of radiation and hence is to be avoided as far as possible. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing indentation test either on outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ condition. Considering this, a remotely operable hydraulic In situProperty Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed in house. The tool head of IProMS can be located inside a pressure tube at any axial location under in situ condition and the properties can be estimated from an analysis of the data on load and depth of indentation, recorded during the test. In order to qualify the system, a number of experimental trials have been conducted on spool pieces and specimens prepared from Zr–2.5Nb pressure tube having different mechanical properties. Based on the encouraging results obtained from the qualification trials, IProMS has been used inside a reactor operated

Microstructural studies on steam oxidised Zr-2.5%Nb pressure tube under simulated LOCA condition

International Nuclear Information System (INIS)

Banerjee, Suparna; Sawarn, Tapan K.; Pandit, K.M.; Anantharaman, S.; Srivastava, D.; Sah, D.N.

2013-03-01

Study of the microstructural evolution of Zr-2.5%Nb pressure tube material of Indian Pressurized Heavy Water Reactors (PHWRs) due to steam oxidation at high temperature (in the range 500-1050°C) was carried out on pressure tube coupons. Hydrogen pick up was less than 55 ppm in the samples oxidized at temperatures up to 850°C but high (250-400 ppm) in the samples oxidized in the β phase region (900°C and above). The microstructure of the samples oxidized above the α-Zr/β-Zr transition temperature showed from the surface inwards sequentially the presence of an oxide layer, an underlying oxygen stabilized α-Zr layer and a prior β-Zr phase containing hydride precipitates. An increase in the hardness was observed near the oxide-metal interface in the coupons oxidized above 900°C, due to formation of oxygen stabilized α-Zr layer. Higher hardness was also observed in the base metal in the samples oxidized at 1000 and 1050°C (author)

X-ray diffraction residual stress measurement in the rolled-joint zone of Zr - 2.5 % Nb pressure tube

International Nuclear Information System (INIS)

Dinu, A.; Nedelcu, L.

1995-01-01

The in-service experience of Zr - 2.5 % Nb pressure tubes in CANDU-type nuclear reactors has demonstrated very good performance over a long period of time. However, analyses done by AECL specialists on most failure cases, showed that a big percentage of defects are manufacturing defects, which appear mostly at the beginning of the rolled-joint zone. It has been observed that a correct rolling ensures an acceptable distribution of residual stress, but an incorrect one leads to an accumulation of big values of residual stress. This determines a preferential radial orientation of hydrides, which during operation in the reactor can produce DHC. To ensure a suitable performance of the Zr - 2.5 % Nb pressure tubes in the CANDU reactor, it is very important to have a correct rolling as mentioned in the procedure. This work presents a methodology for the measurement of the stressing state in the surfaces layers of the rolled-joint zone. The X-ray diffraction method can also be used for establishing the residual stress distribution across the tub wall, in order to ensure a good performance at Cernavoda nuclear plant. The results obtained for the investigated tube have led to the conclusion that the rolling process was correctly applied in this case, the values obtained for the residual stress being in good agreement with those accepted in literature. (Author) 2 Figs., 2 Tabs

Estimation of fracture toughness of Zr 2.5% Nb pressure tube of Pressurised Heavy Water Reactor using cyclic ball indentation technique

Energy Technology Data Exchange (ETDEWEB)

Chatterjee, S., E-mail: subrata@barc.gov.in; Panwar, Sanjay; Madhusoodanan, K.; Rama Rao, A.

2016-08-15

Highlights: • Measurement of fracture toughness of pressure tube is required for its fitness assessment. • Pressure tube removal from the core consumes large amount of radiation for laboratory test. • A remotely operable In situ Property Measurement System (IProMS) has been designed in house. • Conventional and IProMS tests conducted on pressure tube spool pieces having different mechanical properties. • Correlation has been established between the conventional and IProMS estimated fracture properties. - Abstract: In Pressurised Heavy Water Reactors (PHWRs) fuel bundles are located inside horizontal pressure tubes made up of Zr 2.5 wt% Nb alloy. Pressure tubes undergo degradation during its service life due to high pressure, high temperature and radiation environment. Measurement of mechanical properties of degraded pressure tubes is important for assessing their fitness for further operation. Presently as per safety guidelines imposed by the regulatory body, a few pre-decided pressure tubes are removed from the reactor core at regular intervals during the planned reactor shut down to carry out post irradiation examination (PIE) in a laboratory which consumes lots of man-rem and imposes economic penalties. Hence a system is indeed felt necessary which can carry out experimental trials for measurement of mechanical properties of pressure tubes under in situ conditions. The only way to accomplish this important objective is to develop a system based on an in situ measurement technique. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing an indentation test either on the outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ conditions. Considering the importance of such measurements, an In situ Property

Corrosion kinetics at high pressure and temperature of Zr-2.5 Nb with different heat treatments

International Nuclear Information System (INIS)

Jaime Solis, F.; Bordoni, Roberto; Olmedo, Ana M.; Villegas, Marina; Miyagusuku, Marcela

2003-01-01

The corrosion behaviour of Zr-2.5 Nb pressure tube (PT) specimens, with ageing treatments at 400 and 500 C degrees for different times, was studied. The results were analyzed using the corrosion behavior of Zr-20 Nb and Zr-1 Nb samples heat treated during 1 hour at 850 C degrees, cooled in air and aged at the same temperature and times than the PT specimens. The comparison between the corrosion behaviour of Zr-1 Nb and Zr-20 Nb aged coupons with the aged pressure tube specimens, together with the metal/oxide interface morphology of Zr-2.5 Nb specimens, suggest that the increase in the corrosion resistance in the latter coupons is associated with the decomposition of the β-Zr phase. There is also a contribution of α-Zr phase when the ageing temperatures are high enough or the ageing times are long enough, due to a decrease in the Nb content of this phase. This last contribution is associated with an increase in the corrosion resistance of the central zone of pressure tube in the reactor. (author)

Domain sizing diffraction and deformation degree of Zr-2.5%Nb pressure pipes of national production

International Nuclear Information System (INIS)

Banchik, A D; Buoli, D C; Flores, A; Vizcaino, P

2012-01-01

The Life Extension Program for the PHWR Embalse power reactor requires the replacement of the original Zr-2, 5% Nb pressure tubes by a new set of Zr-2,5%Nb pressure tubes. It was also decided to make the new set de pressure tubes locally from imported extruded tubes. As the mechanical properties of pressure tubes in operation depends on the microstructural deformation it is necessary to add to the normal macroscopic quality control same microstructural studies, specially .during the development stage. In the present work x-ray diffraction techniques are applied to determine the magnitude of the micro-structural deformation of three pressure tubes obtained Zr-2,5% Nb extruded tubes that were processes following designer specifications. The pressure tubes were obtained by two step of plastic deformation in a HPTR 60-120 tube rolling machine without intermediate annealing and a final thermal treatment. The line width at half height (FWHM) diffraction lines diagrams are drawn in a Williamson - Hall plot to determine the size of the coherent diffraction domains and estimate the degree of micro-structural deformation. To evaluate the possible effects of texture and/or anisotropy the X-ray measurement were made at the three principal directions of the processed tubes. In summary, the three pressure tubes obtained follow the same average Williamson-Hall line and similar dispersion with respect to that line. That result implies an acceptable homogeneity between the micro-deformation of these pressure tubes. As then have been chosen at random from a lot of preceded pressure tubes, it is possible to predict similar conclusions for the entire batch (author)

Microstructure and textural characterization of hot extruded Zr-2.5Nb alloy PHWR pressure tube fabricated by various ingot processing route

International Nuclear Information System (INIS)

Vaibhaw, Kumar; Jha, S.K.; Saibaba, N.; Neogy, S.; Mani Krishna, K.V.; Srivastava, D.; Dey, G.K.

2011-01-01

Zr-2.5 Nb alloys finds its applications as a pressure tube component in pressure tube type thermal reactors such as PHWRs and RBMK due to properties attributed such as low neutron absorption cross section, high temperature strength and corrosion resistance etc. Manufacturing of this life time components involves series of thermo-mechanical processes of hot working and cold working with intermediate annealing. The life time of Pressure tube are limited due to their diametral creep properties which is governed by metallurgical characteristics such as texture, microstructure dislocation density etc. The primary breakdown of cast structure in Vacuum Arc Melted ingot can be effected by either hot extrusion or forging in single or multiple stages before final hot extrusion step into the blank for manufacturing of seamless pressure tube. Elevated temperature deformation carried out in hot working above the recrystallization temperature would enable impositions of large strains in single step. This deformation causes a significant change in the microstructure of the material and depends on process parameters such as extrusion ratio, temperature and strain rate. Basic microstructure developed at this deformation stage has significant bearing on the final properties of the material fabricated with subsequent cold working steps. The major texture in α+β Zr-2.5 Nb alloy is established during final extrusion to blank which does not change significantly during subsequent cold pilgering. However, microstructure is modified significantly in subsequent cold working which can be effected by cold pilgering or cold drawing in single or multiple steps. Present paper brings out the various ingot processing routes using forging and or extrusion followed for fabrication of pressure tubes. The development of texture and microstructures has been discussed at the blank stage from these processing routes and also with respect to varying extrusion variable such as extrusion ratio

Influence of hydrogen content on fracture toughness of CWSR Zr-2.5Nb pressure tube alloy

Science.gov (United States)

Singh, R. N.; Bind, A. K.; Srinivasan, N. S.; Ståhle, P.

2013-01-01

In this work, influence of hydrogen and temperature on the fracture toughness parameters of unirradiated, cold worked and stress relieved (CWSR) Zr-2.5Nb pressure tube alloys used in Indian Pressurized Heavy Water Reactor is reported. The fracture toughness tests were carried out using 17 mm width curved compact tension specimens machined from gaseously hydrogen charged tube-sections. Metallography of the samples revealed that hydrides were predominantly oriented along axial-circumferential plane of the tube. Fracture toughness tests were carried out in the temperature range of 30-300 °C as per ASTM standard E-1820-06, with the crack length measured using direct current potential drop (DCPD) technique. The fracture toughness parameters (JQ, JMax and dJ/da), were determined. The critical crack length (CCL) for catastrophic failure was determined using a numerical method. It was observed that for a given test temperature, the fracture toughness parameters representing crack initiation (JQ) and crack propagation (JMax, and dJ/da) is practically unaffected by hydrogen content. Also, for given hydrogen content, all the aforementioned fracture toughness parameters increased with temperature to a saturation value.

Fabrication of Zr-2.5Nb pressure tubes to minimise the harmful effects of trace elements

International Nuclear Information System (INIS)

Theaker, J.R.; Coleman, C.E.; Davis, L.; Graham, R.A.

1994-03-01

Trace elements can reduce the fracture resistance of Zr-2.5Nb pressure tubes. The effects of hydrogen as hydrides and oxygen as an alloy-strengthening agent are well known, but the contribution of carbon, phosphorus, chlorine and segregated oxygen has only recently been recognized. Carbides and phosphides are brittle particles, while chlorine segregates to form planes of weakness that produce fissures is associated with low toughness. With long hold times in the (α + β) region, oxygen partitions in the α-grains; such grains are hard, and if they survive fabrication may reduce the toughness of the finished tube. Through a co-operative program involving AECL and manufacturers, a series of manufacturing innovations and controls has been introduced that minimize these harmful effects. Hydrogen is present in Zr sponge as water, can be absorbed at each stage of tube fabrication, and needs to be carefully controlled, particularly during ingot breakdown and subsequent forging. Hydrogen concentrations in finished tubes have been reduced by a factor of three through the optimization of manufacturing processes and the implementation of new technology. Multiple vacuum arc melting, use of selected raw materials and intermediate ingot surface conditioning have resulted in much improved fracture toughness through the reduction of chlorine and phosphorus concentrations. Optimum distribution of oxygen may be achieved through changes to the extrusion process cycle. An understanding of the Zr-2.5Nb-C phase diagram, particularly the solubility of carbon at low concentrations, has resulted in the specification of a lower carbon concentration. (author). 12 refs., 6 tabs., 10 figs

β-85 wt % Nb precipitates: the effect on in-reactor diametric creep of pressure tubes

International Nuclear Information System (INIS)

Sarce, Alicia

2006-01-01

By linking the microstructure evolution of an α-Zr crystal with the macroscopic behaviour, the deformation of an in-service reactor pressure tube is calculated. Microstructure evolution is considered through rate theory modelling of the interaction between point defects and sinks. Different densities of β-85 wt % Nb precipitates are proposed to be distributed inside the α grains and act as point defect sinks, doing a screening effect on the grain boundaries. From the interatomic pair potential which is used to describe the material, positive tangential deformation rates (on the other hand negatives) are obtained when these densities in the c-crystal direction are bigger than a minimum value. (author) [es

Annular gap measurement between pressure tube and calandria tube by eddy current technique

International Nuclear Information System (INIS)

Bhole, V.M.; Rastogi, P.K.; Kulkarni, P.G.

1992-01-01

In pressurised heavy water reactor (PHWR) major distinguishing feature is that there are number of identical fuel channels in the reactor core. Each channel consists of pressure tube of Zr-2.5 Nb or zircaloy-2 through which high temperature, high pressure primary coolant is passing. The pressure tube contains fuel. Surrounding the pressure tube there is low pressure, cool heavy water (moderator). The moderator is thermally separated from coolant by the tube which is nominally concentric with pressure tube called calandria tube. There are four garter springs in the annular gap between pressure tube and calandria tube. During the life of the reactor there are number of factors by which the pressure tube sags, most important factors are irradiation creep, thermal creep, fuel load etc. Because of the sag of pressure tube it can touch the calandria tube resulting in formation of cold spot. This leads to hydrogen concentration at that spot by which the material at that place becomes brittle and can lead to catastrophic failure of pressure tube. There is no useful access for measurement of annular gap either through the gas annular space or from exterior of calandria tube. So the annular gap was measured from inside surface of pressure tube which is accessible. Eddy current technique was used for finding the gap. The paper describe the details of split coil design of bobbin probe, selection of operating point on normalised impedance diagram by choosing frequency. Experimental results on full scale mock up, and actual gap measurement in reactor channel, are also given. (author). 7 figs

Highlights of the metallurgical behaviour of CANDU pressure tubes

International Nuclear Information System (INIS)

Price, E.G.

1984-10-01

This paper is an overview of the service induced metallurgical changes that take place in Zircaloy-2 and Zr-2.5 wt. percent Nb pressure tubes in CANDU reactors. It incorporates the findings of an evaluation program, that followed a significant pressure tube failure at Ontario Hydro's Pickering Nuclear Generating Station, and also provides valid reasons for continued confidence in the current CANDU design

Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

International Nuclear Information System (INIS)

Krishnan, S.; Bhasin, V.; Mahajan, S.C.

1997-01-01

Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

Energy Technology Data Exchange (ETDEWEB)

Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

1997-04-01

Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

Development of rolled joints for zirconium-2.5 wt % niobium pressure tubes

International Nuclear Information System (INIS)

Madhusoodanan, K.; Sinha, R.K.; Samuel, K.A.; Joeman, V.

1992-01-01

Due to its higher strength and lower deuterium pick-up rate, as compared to the existing cold worked zircaloy-2 material, cold worked zirconium-2.5 wt% niobium (Zr-2.5%Nb) alloy is to be used as the pressure tube material in all forthcoming Indian PHWRs starting with KAPP-2. These pressure tubes, which carry the fuel bundles are to be joined to the S.S 403 end-fittings through rolled joints. Since the new pressure tubes have a lower wall thickness and higher room temperature yield stress, than zircaloy-2 tubes the design parameters of the rolled joint had to be developed afresh. Further, since Zr-2.5%Nb is susceptible to delayed hydride cracking, it is necessary to limit the residual stress near the rolled joint to a minimum. Since the high residual stress is due to the initial assembly clearance between the pressure tube and end-fitting, a modified rolled joint had to be developed, referred to as zero clearance rolled joint. This paper provides details of the work carried out at Reactor Engineering Division of Bhabha Atomic Research Centre, Bombay towards the development of the design of the rolled joint as well as the tooling and procedures required for achieving zero-clearance fit-ups at site. The requirements to be met by the Zr-2.5% Nb pressure tubes for achieving acceptable rolled joints are highlighted. (author). 5 refs., 6 figs., 3 tabs

On the anisotropy of in-reactor creep of Zr-2.5Nb tubes

International Nuclear Information System (INIS)

Causey, A.R.; Holt, R.A.

1993-06-01

Creep specimens made from cold-worked Zr-2.5Nb tubes, fabricated with two different microstructures and crystallographic textures, were irradiated in the Osiris reactor in France in a fast-neutron flux of about 1.8 x 10 18 n.m -2 .s -1 , E > MeV, at 553 and 585 K. The hoop stresses from internal Fluences, up to 4 x 10 25 n.m -2 , more than double those achieved an any other creep test on cold-worked Zr-2.5Nb in which both axial and transverse strain were measured. Creep rates were obtained from strain versus fluence plots, and creep compliances were obtained from plots of the strain rates against hoop stress for each material at each temperature. The ratio of creep rates at 583 K to those at 553 K was ∼ 1.36, a little higher than that extrapolated from stress relaxation results at temperatures between 523 and 568 K. The ratio of the biaxial creep compliance in the axial direction to that in the transverse directions is different for the two test materials: 0.0 to -0.1 for the fuel sheathing texture and 0.5 to 0.6 for the pressure tube texture. The results were analysed using a self-consistent model developed to account for the contributions to the creep anisotropy of the three microstructure parameters involved and to account for the grain interaction effects. The model, which was normalized to test reactor and power reactor creep data for cold-worked Zr-2.5Nb tubes, predicted the ratio of the creep compliancies to be -0.26 and 0.63, respectively. Thus the creep anisotropy of Zr-2.5Nb tubes with pressure-tube-like crystallographic texture can be adequately predicted. (author). 18 refs., 4 tabs., 13 figs

Development of Zirconium alloys (for pressure tubes)

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Jung, Chung Hwan; Yim, Kyong Soo; Kim, Sung Soo; Baek, Jong Hyuk; Jeong, Yong Hwan; Kim, Kyong Ho; Cho, Hae Dong [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Hwang, S. K.; Kim, M. H. [Inha Univ., Incheon (Korea, Republic of); Kwon, S. I [Korea Univ., Seoul (Korea, Republic of); Kim, I. S. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

1997-09-01

The objective of this research is to set up the basic technologies for the evaluation of pressure tube integrity and to develop improved zirconium alloys to prevent pressure tube failures due to DHC and hydride blister caused by excessive creep-down of pressure tubes. The experimental procedure and facilities for characterization of pressure tubes were developed. The basic research related to a better understanding of the in-reactor performances of pressure tubes leads to noticeable findings for the first time : the microstructural effect on corrosion and hydrogen pick-up behavior of Zr-2.5Nb pressure tubes, texture effect on strength and DHC resistance and enhanced recrystallization by Fe in zirconium alloys and etc. Analytical methodology for the assessment of pressure tubes with surface flaws was set up. A joint research is being under way with AECL to determine the fracture toughness of O-8 at the EOL (End of Life) that had been quadruple melted and was taken out of the Wolsung Unit-1 after 10 year operation. In addition, pressure tube with texture controlled is being made along with VNINM in Russia as a joint project between KAERI and Russia. Finally, we succeeded in developing 4 different kinds of zirconium alloys with better corrosion resistance, low hydrogen pickup fraction and higher creep strength. (author). 121 refs., 65 tabs., 260 figs

Evaluation of oxides formed at high temperatures in Zr-2.5Nb pressure tubing

Energy Technology Data Exchange (ETDEWEB)

Kiran Kumar, N.A.P.; Szpunar, J.A., E-mail: kiraniitkgp@yahoo.com [Univ. of Saskatchewan, Saskatoon, Saskatchewan (Canada)

2012-07-01

The oxidation behavior of Zr-2.5Nb pressure tube samples has been studied at four different temperatures, i.e., 400°, 600°, 800°, and 1000°C. The amount of tetragonal phase is found to decrease with increase of temperature. The oxide texture of (002){sub m} and (111){sub m} type increased with the temperature from 400°C to 600°C, however at temperatures above 600°C the texture strength seems to diminish and the oxide layer becomes structurally unstable. Further, the impedance response is found to be dependent on the microstructure of the oxide film. For the sample oxidized at 400°C, Electrochemical Impedance Spectroscopy (EIS) spectra exhibited a two-time constant behavior, showing the formation of two-layer oxide film on the Zr-2.5Nb alloy, which correspond to a porous outer oxide and a barrier inner oxide, respectively. In addition, the samples were oxidized at constant temperature of 600°C with varying oxidation time. The observation shows that the oxide is more protective in the early stage of oxide growth. However, further growth of oxide film has resulted in degeneration of its protective character. (author)

Effect of heat treatments and pre-irradiation on the corrosion at high pressure and temperature of pressure tube Zr-2.5Nb

International Nuclear Information System (INIS)

Olmedo, A.M; Bordoni, R

2006-01-01

Zr and its alloys are widely used in the nuclear industry. The pressure tubes (PT) of CANDU reactors are made using the alloy Zr-2.5Nb because of its high mechanical resistance and its good behavior under aqueous corrosion at high pressure and temperature. The resulting microstructure after the production process consists of elongated grains of Zr-α with hcp structure (≤ 1%Nb) , approximately 0.3-0.5 μ thick, surrounded by a metastable β Zr bcc phase rich in Nb (≅20%). Temperatures of less than 600 C produce an evolution of the β Zr phase towards an equilibrium structure consisting of β Nb (≅85% of Nb) and α Zr phases. The conditions for the evolution of this phase may involve a Nb-poor intermediate phase called ω and the enrichment of Nb for the rest. The radiation, particularly the flow of rapid neutrons (E ≥1Mev), modifies the microstructure of the material inducing β-Nb precipitates in the grains of the α-Zr phase. The temperature and radiation induced changes in the microstructure modify the resistance to the corrosion of the Zr-2.5Nb of the PT. This work studies the behavior to corrosion of Zr-2.5Nb coupons of (PT) with aging heat treatments (HT) of (PT) at 400 o C for 72 and 1000 hours and at 500 o C for 2.6 and 10 hours. The results indicated that the corrosion speed of the test pieces with HT is less than for those without HT and that the decrement of the corrosion speed is much more pronounced in the first 2.6 h of aging at 500 o C and in the first 72 h at 400 o C. Also the greater the duration of the HT, the better the behavior to corrosion. This decreased corrosion speed is associated with the modifications produced in the microstructure of the PT material due to the effect of the HT, which cause the decomposition of the βZr phase and lead to a microstructure that is much closer to one of equilibrium. That is, the closer the material's microstructure is to equilibrium the greater will be the resistance to its corrosion. This effect

Effect of error in crack length measurement on maximum load fracture toughness of Zr-2.5Nb pressure tube material

International Nuclear Information System (INIS)

Bind, A.K.; Sunil, Saurav; Singh, R.N.; Chakravartty, J.K.

2016-03-01

Recently it was found that maximum load toughness (J max ) for Zr-2.5Nb pressure tube material was practically unaffected by error in Δ a . To check the sensitivity of the J max to error in Δ a measurement, the J max was calculated assuming no crack growth up to the maximum load (P max ) for as received and hydrogen charged Zr-2.5Nb pressure tube material. For load up to the P max , the J values calculated assuming no crack growth (J NC ) were slightly higher than that calculated based on Δ a measured using DCPD technique (JDCPD). In general, error in the J calculation found to be increased exponentially with Δ a . The error in J max calculation was increased with an increase in Δ a and a decrease in J max . Based on deformation theory of J, an analytic criterion was developed to check the insensitivity of the J max to error in Δ a . There was very good linear relation was found between the J max calculated based on Δ a measured using DCPD technique and the J max calculated assuming no crack growth. This relation will be very useful to calculate J max without measuring the crack growth during fracture test especially for irradiated material. (author)

Study of microstructure, texture and mechanical properties of Zr–2.5Nb alloy pressure tubes fabricated with different processing routes

International Nuclear Information System (INIS)

Saibaba, N.; Vaibhaw, Kumar; Neogy, S.; Mani Krishna, K.V.; Jha, S.K.; Phani Babu, C.; Ramana Rao, S.V.; Srivastava, D.; Dey, G.K.

2013-01-01

Different fabrication trials involving the variation in three important stages of Zr–2.5Nb pressure tube were undertaken. The variations were with respect to the mode of breaking the cast structure of the ingot (forging vs extrusion), the hot extrusion ratio and the number of subsequent cold work stages to produce the finished tube. It was observed that the forging process resulted in superior performance in breaking the cast structure. Higher extrusion ratios resulted in more favorable texture and microstructure. More continuity of the beta phase was observed in the final microstructure for the route involving the single cold work step subsequent to hot extrusion

An improved model to predict nonuniform deformation of Zr-2.5 Nb pressure tubes

International Nuclear Information System (INIS)

Lei, Q.M.; Fan, H.Z.

1997-01-01

Present circular pressure-tube ballooning models in most fuel channel codes assume that the pressure tube remains circular during ballooning. This model provides adequate predictions of pressure-tube ballooning behaviour when the pressure tube (PT) and the calandria tube (CT) are concentric and when a small (<100 degrees C) top-to-bottom circumferential temperature gradient is present on the pressure tube. However, nonconcentric ballooning is expected to occur under certain postulated CANDU (CANada Deuterium Uranium) accident conditions. This circular geometry assumption prevents the model from accurately predicting nonuniform pressure-tube straining and local PT/CT contact when the pressure tube is subjected to a large circumferential temperature gradient and consequently deforms in a noncircular pattern. This paper describes an improved model that predicts noncircular pressure-tube deformation. Use of this model (once fully validated) will reduce uncertainties in the prediction of pressure-tube ballooning during a postulated loss-of-coolant accident (LOCA) in a CANDU reactor. The noncircular deformation model considers a ring or cross-section of a pressure tube with unit axial length to calculate deformation in the radial and circumferential directions. The model keeps track of the thinning of the pressure-tube wall as well as the shape deviation from a reference circle. Such deviation is expressed in a cosine Fourier series for the lateral symmetry case. The coefficients of the series for the first m terms are calculated by solving a set of algebraic equations at each time step. The model also takes into account the effects of pressure-tube sag or bow on ballooning, using an input value of the offset distance between the centre of the calandria tube and the initial centre of the pressure tube for determining the position radius of the pressure tube. One significant improvement realized in using the noncircular deformation model is a more accurate prediction in

Structural integrity evaluations of CANDU pressure tubes

International Nuclear Information System (INIS)

Radu, Vasile

2003-01-01

The core of a CANDU-6 pressurized heavy water reactor consists of some hundred horizontal pressure tubes that are manufactured from a Zr-2.5%Nb alloy and which contain the fuel bundles. These tubes are susceptible to a damaging phenomenon known as Delayed Hydride Cracking (DHC). The Zr-2.5%Nb alloy is susceptible to DHC phenomenon when there is diffusion of hydrogen atoms to a service-induced flaws, followed by the hydride platelets formation on the certain crystallographic planes in the matrix material. Finally, the development of hydride regions at the flaw-tip will happened. These hydride regions are able to fracture under stress-temperature conditions (DHC initiation) and the cracks can extend and grow by DHC mechanism. Some studies have been focused on the potential to initiate DHC at the blunt flaws in a CANDU reactor pressure tube and a methodology for structural integrity evaluation was developed. The methodology based on the Failure Assessment Diagrams (FAD's) consists in an integrated graphical plot, where the fracture failure and plastic collapse are simultaneously evaluated by means of two non-dimensional variables (K r and L r ). These two variables represent the ratio of the applied value of either stress or stress intensity factor and the resistance parameter of corresponding magnitude (yield stress or fracture toughness, respectively). Once the plotting plane is determined by the variables K r and L r , the procedure defines a critical failure line that establishes the safe area. The paper will demonstrate the possibility to perform structural integrity evaluations by means of Failure Assessment Diagrams for flaws occurring in CANDU pressure tubes. (author)

Sample summary report for ARG 1 pressure tube sample

International Nuclear Information System (INIS)

Belinco, C.

2006-01-01

The ARG 1 sample is made from an un-irradiated Zr-2.5% Nb pressure tube. The sample has 103.4 mm ID, 112 mm OD and approximately 500 mm length. A punch mark was made very close to one end of the sample. The punch mark indicates the 12 O'clock position and also identifies the face of the tube for making all the measurements. ARG 1 sample contains flaws on ID and OD surface. There was no intentional flaw within the wall of the pressure tube sample. Once the flaws are machined the pressure tube sample was covered from outside to hide the OD flaws. Approximately 50 mm length of pressure tube was left open at both the ends to facilitate the holding of sample in the fixtures for inspection. No flaw was machined in this zone of 50 mm on either end of the pressure tube sample. A total of 20 flaws were machined in ARG 1 sample. Out of these, 16 flaws were on the OD surface and the remaining 4 on the ID surface of the pressure tube. The flaws were characterized in to various groups like axial flaws, circumferential flaws, etc

Delayed hydrogen cracking test design for pressure tubes

International Nuclear Information System (INIS)

Haddad, Roberto; Loberse, Antonio N.; Yawny, Alejandro A.; Riquelme, Pablo

1999-01-01

CANDU nuclear power stations pressure tubes of alloy Zr-2,5 % Nb present a cracking phenomenon known as delayed hydrogen cracking (DHC). This is a brittle fracture of zirconium hydrides that are developed by hydrogen due to aqueous corrosion on the metal surface. This hydrogen diffuses to the crack tip where brittle zirconium hydrides develops and promotes the crack propagation. A direct current potential decay (DCPD) technique has been developed to measure crack propagation rates on compact test (CT) samples machined from a non irradiated pressure tube. Those test samples were hydrogen charged by cathodic polarization in an acid solution and then pre cracked in a fatigue machine. This technique proved to be useful to measure crack propagation rates with at least 1% accuracy for DHC in pressure tubes. (author)

A study on the delayed hydride cracking mechanism in cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and zircaloy-2 pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Choi, Kwang Sik

1992-02-15

Cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and Zircaloy-2 pressure tubes were hydrided to the hydrogen concentration of 68 ppm, 49 ppm and 242-411 ppm, respectively, and compact tension specimens were machined from the hydrided materials. The crack growth rate by delayed hydride cracking was measured by potential drop method at various temperatures on the above mentioned three types of specimens. The activation energy obtained were 43 KJ/mol for cold worked Zr-2.5Nb and 37 KJ/mol for heat treated Zr-2.5Nb, which were in good agreements with that of Coleman (1977), while they were lower than the activation energy of 65.5 KJ/mol obtained by Simpson-puls (1979) and 71.5 KJ/mol by Ambler (1984). The DHC growth rate in Zircaloy-2 were about one fifth of that of Zr-2.5Nb, which is due to the texture and material strength effects. Striations which indicate stepwise DHC growth were observed at fracture surface by scanning electron microscope and unsymmetric crack tunnellings were also observed, which seems to be due to the difference in hydrogen diffusion rate caused by the difference in stress fields between inner and outer surface. The comparison of test results with the DHC growth rate calculated by Simpson-puls model showed good agreement at high temperatures, whereas at the lower temperatures the crack growth rates were 2.5 times higher than the calculated values.

A study on the delayed hydride cracking mechanism in cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and zircaloy-2 pressure tubes

International Nuclear Information System (INIS)

Choi, Kwang Sik

1992-02-01

Cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and Zircaloy-2 pressure tubes were hydrided to the hydrogen concentration of 68 ppm, 49 ppm and 242-411 ppm, respectively, and compact tension specimens were machined from the hydrided materials. The crack growth rate by delayed hydride cracking was measured by potential drop method at various temperatures on the above mentioned three types of specimens. The activation energy obtained were 43 KJ/mol for cold worked Zr-2.5Nb and 37 KJ/mol for heat treated Zr-2.5Nb, which were in good agreements with that of Coleman (1977), while they were lower than the activation energy of 65.5 KJ/mol obtained by Simpson-puls (1979) and 71.5 KJ/mol by Ambler (1984). The DHC growth rate in Zircaloy-2 were about one fifth of that of Zr-2.5Nb, which is due to the texture and material strength effects. Striations which indicate stepwise DHC growth were observed at fracture surface by scanning electron microscope and unsymmetric crack tunnellings were also observed, which seems to be due to the difference in hydrogen diffusion rate caused by the difference in stress fields between inner and outer surface. The comparison of test results with the DHC growth rate calculated by Simpson-puls model showed good agreement at high temperatures, whereas at the lower temperatures the crack growth rates were 2.5 times higher than the calculated values

Deformation behavior of irradiated Zr-2.5Nb pressure tube material

International Nuclear Information System (INIS)

Himbeault, D.D.; Chow, C.K.; Puls, M.P.

1994-01-01

A study of the deformation behavior of irradiated highly textured Zr-2.5Nb pressure tube material in the temperature range of 30 degree C to 300 degree C was undertaken to understand better the mechanism for the deterioration of the fracture toughness with neutron irradiation. Strain localization behavior, believed to be a main contributor to reduced toughness, was observed in irradiated transverse tensile specimens at temperature greater than 100 degree C. The strain localization behavior was found to occur by the cooperative twinning of the highly textured grains of the material, resulting in a local softening of the material, where the flow than localizes. It is believed that the effect of the irradiation is to favor twinning at the expense of slip in the early stages of deformation. This effect becomes more pronounced at higher temperature, thus leading to the high-temperature strain localization behavior of the material. A limited amount of dislocation channeling was also observed; however, it is not considered to have a major role in the strain localization behavior of the material. Contrary to previous reports on irradiated zirconium alloys, static strain aging is observed in the irradiated material in the temperature range of 150 degree C to 300 degree C

Flow behaviour of autoclaved, 20% cold worked, Zr-2.5Nb alloy pressure tube material in the temperature range of room temperature to 800 deg. C

International Nuclear Information System (INIS)

Dureja, A.K.; Sinha, S.K.; Srivastava, Ankit; Sinha, R.K.; Chakravartty, J.K.; Seshu, P.; Pawaskar, D.N.

2011-01-01

Pressure tube material of Indian Heavy Water Reactors is 20% cold-worked and stress relieved Zr-2.5Nb alloy. Inherent variability in the process parameters during the fabrication stages of pressure tube and also along the length of component have their effect on micro-structural and texture properties of the material, which in turn affect its strength parameters (yield strength and ultimate tensile strength) and flow characteristics. Data of tensile tests carried out in the temperature range from room temperature to 800 deg. C using the samples taken out from a single pressure tube have been used to develop correlations for characterizing the strength parameters' variation as a function of axial location along length of the tube and the test temperature. Applicability of Ramberg-Osgood, Holloman and Voce's correlations for defining the post yield behaviour of the material has been investigated. Effect of strain rate change on the deformation behaviour has also been studied.

A statistical approach to the prediction of pressure tube fracture toughness

International Nuclear Information System (INIS)

Pandey, M.D.; Radford, D.D.

2008-01-01

The fracture toughness of the zirconium alloy (Zr-2.5Nb) is an important parameter in determining the flaw tolerance for operation of pressure tubes in a nuclear reactor. Fracture toughness data have been generated by performing rising pressure burst tests on sections of pressure tubes removed from operating reactors. The test data were used to generate a lower-bound fracture toughness curve, which is used in defining the operational limits of pressure tubes. The paper presents a comprehensive statistical analysis of burst test data and develops a multivariate statistical model to relate toughness with material chemistry, mechanical properties, and operational history. The proposed model can be useful in predicting fracture toughness of specific in-service pressure tubes, thereby minimizing conservatism associated with a generic lower-bound approach

Inert medium (helium) irradiation testing of pressure tube samples

International Nuclear Information System (INIS)

Ancuta, M.; Radu, V.; Stefan, V.; Preda, M.

2001-01-01

Irradiation tests currently performed in C-5 capsule aim at obtaining data and information concerning behavior to irradiation of pressure tubes of CANDU type fuel channel, to evidence the factors limiting operation life span. A calculation code for analysis and prediction of pressure tube behavior should be based upon periodical inspection results, post irradiation examination of the removed from reactor pressure tubes as well as on the experimental results obtained with materials subjected to irradiation conditions identical with the operational ones. Mechanical behavior analysis should focus both complex thermal-mechanical type stresses and mechanical properties alteration under irradiation. The experimental results should be applied: - to evaluate the irradiation effects upon mechanical properties of Zr-2.5% Nb exposed to fluences up to 10 21 n·cm -2 ; - to gather data concerning the real stress / real deformation characteristic from which characteristic quantities can be deduced as, for instance, elasticity modulus, plasticity modulus, exponent of stress term in the Tsu-Berteles relation, to be used within the CANTUP simulation code describing pressure tube behavior, currently developed at INR Pitesti; - to develop prediction methods of pressure tube behavior and merging with in-service inspection procedure in order to forecast the life span and the proper timing for replacement before major failures occur. The samples irradiated in C-5 capsule were extracted from the ends of Zr-2.5% Nb pressure tubes resulting from Cernavoda NPP Unit 1. The samples for tensile tests were extracted on longitudinal and transversal directions of the pressure tube. The tests were carried out under following conditions: - test environment temperature, 260 - 280 deg.C; - testing medium, helium at 1 - 6 b pressure; - neutron flux (E n > 1 MeV), 1 - 2 · 10 13 ncm -2 s -1 ; - neutron fluence (E n > 1 MeV), 4 · 10 20 ncm -2 . The following characteristics were obtained from tensile

Suppressing hydrogen ingress during aqueous corrosion of CANDU Zr-2.5 Nb pressure tube material

International Nuclear Information System (INIS)

Elmoselhi, M.B.; Donner, A.; Brennenstuhl, A.; Warr, B.D.; Ellis, P.J.; Evans, D.W.

2002-01-01

As a result of their special properties, including low neutron cross-section and intrinsic corrosion resistance, Zr alloys are used in the fabrication of nuclear core components, particularly fuel cladding (in most reactor types) and also Zr-2.5 Nb pressure tubes in CANDU trademark (Canada Deuterium Uranium) reactors. Corrosion and H uptake during service can limit the life of these components. Therefore, remedial action may be appropriate to slow the H uptake rate and prolong the working life of these reactor components. This work has explored the possibility of reducing H uptake in pressure tube material by incorporating an inhibiting agent into the corrosion environment. Two approaches have been tested, depositing a thin metallic film on the initial oxide surface and adding an inhibiting agent to the solution. The latter approach appears more practical. Screening experiments were conducted in short-term (∝30 day) exposures in high temperature (340 C) aqueous out-reactor environments, simulating the CANDU trademark heat transport coolant with various chemistries. Compounds tested included aluminum acetate, aluminum nitrate, lithium nitrate, rhodium nitrate and yttrium nitrate. Comparison of results from the aluminum nitrate additives and aluminum acetate additives suggests that the nitrate anion is the effective ingredient for H ingress inhibition. The nitrate anion appears to reduce the rate of H ingress regardless of the associated cation. However, each cation appears to affect the rate of corrosion differently. These cations were found to be incorporated in the oxide film. (authors)

High-temperature transient creep properties of CANDU pressure tubes

International Nuclear Information System (INIS)

Fong, R.W.L.; Chow, C.K.

2002-06-01

During a hypothetical large break loss-of-coolant accident (LOCA), the coolant flow would be reduced in some fuel channels and would stagnate and cause the fuel temperature to rise and overheat the pressure tube. The overheated pressure tube could balloon (creep radially) into contact with its moderator-cooled calandria tube. Upon contact, the stored thermal energy in the pressure tube is transferred to the calandria tube and into the moderator, which acts as a heat sink. For safety analyses, the modelling of fuel channel deformation behaviour during a large LOCA requires a sound knowledge of the high-temperature creep properties of Zr-2.5Nb pressure tubes. To this extent, a ballooning model to predict pressure-tube deformation was developed by Shewfelt et al., based on creep equations derived using uniaxial tensile specimens. It has been recognized, however, that there is an inherent variability in the high-temperature creep properties of CANDU pressure tubes. The variability, can be due to different tube-manufacturing practices, variations in chemical compositions, and changes in microstructure induced by irradiation during service in the reactor. It is important to quantify the variability of high-temperature creep properties so that accurate predictions on pressure-tube creep behaviour can be made. This paper summarizes recent data obtained from high-temperature uniaxial creep tests performed on specimens taken from both unirradiated (offcut) and irradiated pressure tubes, suggesting that the variability is attributed mainly to the initial differences in microstructure (grain size, shape and preferred orientation) and also from tube-to-tube variations in chemical composition, rather than due to irradiation exposure. These data will provide safety analysts with the means to quantify the uncertainties in the prediction of pressure-tube contact temperatures during a postulated large break LOCA. (author)

Ballooning of CANDU pressure tube in local thermal transients

International Nuclear Information System (INIS)

Mihalache, Maria; Ionescu, Viorel

2008-01-01

In certain LOCA scenarios for the CANDU fuel channel, the ballooning of the pressure tube and contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator takes place through the contact area. If the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. In INR-Pitesti the DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after occurrence of the contact between the two tubes. The code contains few models: thermal creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This paper gives a DELOCA code description and the fuel channel behaviour analysis, in transient temperature conditions of the pressure tube, using the materials properties, time and temperature dependencies of these properties as obtained in the different laboratories of the world and in the INR - Pitesti in the last years. DELOCA computer code simulated the fuel channel response to the constant heating rates of inside pressure tube surface. The paper presents contact temperature and time dependencies on the heating rate, and the appropriate fitting functions. (authors)

Operating envelope to minimize probability of fractures in Zircaloy-2 pressure tubes

International Nuclear Information System (INIS)

Azer, N.; Wong, H.

1994-01-01

The failure mode of primary concern with Candu pressure tubes is fast fracture of a through-wall axial crack, resulting from delayed hydride crack growth. The application of operating envelopes is demonstrated to minimize the probability of fracture in Zircaloy-2 pressure tubes based on Zr-2.5%Nb pressure tube experience. The technical basis for the development of the operating envelopes is also summarized. The operating envelope represents an area on the pressure versus temperature diagram within which the reactor may be operated without undue concern for pressure tube fracture. The envelopes presented address both normal operating conditions and the condition where a pressure tube leak has been detected. The examples in this paper are prepared to illustrate the methodology, and are not intended to be directly applicable to the operation of any specific reactor. The application of operating envelopes to minimized the probability of fracture in 80 mm diameter Zircaloy-2 pressure tubes has been discussed. Both normal operating and leaking pressure tube conditions have been considered. 3 refs., 4 figs

Fracture Toughness Round Robin Test International in pressure tube materials

International Nuclear Information System (INIS)

Villagarcia, M.P.; Liendo, M.F.

1993-01-01

Part of the pressure tubes surveillance program of CANDU type reactors is to determine the fracture toughness using a special fracture specimen and test procedure. Atomic Energy of Canada Limited decided to hold a Round Robin Test International and 9 laboratories participated worldwide in which several pressure tube materials were selected: Zircaloy-2, Zr-2.5%Nb cold worked and Zr-2.5%Nb heat treated. The small specimens used held back the thickness and curvature of the tube. J-R curves at room temperature were obtained and the crack extension values were determined by electrical potential drop techniques. These values were compared with results generated from other laboratories and a bid scatter was founded. It could be due to slight variations in the test method or inhomogeneity of the materials and a statistical study must be done to see if there is any pattern. The next step for the Round Robin Test would be to make some modifications in the test method in order to reduce the scatter. (Author)

Examination of parameters affecting overload fracture behavior of flaw-tip hydrides in Zr-2.5Nb pressure tubes in Candu reactors

International Nuclear Information System (INIS)

Cui, J.; Shek, G.K.; Wang, Z.R.

2007-01-01

Service-induced flaws in Zr-2.5Nb alloy pressure tubes in Candu (Canada Deuterium Uranium Reactors) nuclear reactors are susceptible to a crack initiation and growth mechanism known as Delayed Hydride Cracking (DHC), which is a repetitive process that involves hydrogen diffusion, hydride precipitation, growth and fracture of a hydride region at the flaw-tip under a constant load. Crack initiation may also occur under another loading condition when the hydride region is subjected to an overload. An overload occurs when the hydride region at the flaw tip is loaded to a stress higher than that at which this region is formed such as when the reactor experiences a transient pressure higher than the normal operating pressure where the hydride region is formed. Flaw disposition requires justification that the hydride region overload will not fracture the hydride region, and initiate DHC. In this work, monotonically increasing load experiments were performed on unirradiated Zr-2.5Nb pressure tube specimens containing simulated debris frets (V-notch) and bearing pad frets (BPF, U-shape notch) to examine overload fracture behavior of flaw-tip hydrides formed under hydride ratcheting conditions. Hydride cracking in the overload tests was detected by the acoustic emission technique and confirmed by post-test metallurgical examination. Test results indicate that the resistance to overload fracture is affected by a number of parameters including hydride formation stress, flaw shape (V-notch vs. BPF) and flaw radius (0.015 mm vs. 0.1 mm). The notch-tip hydride morphologies were examined by optical microscopy and scanning electron microscopy (SEM) which show that they are affected by the hydride formation conditions, resulting in different overload fracture resistance. Finite element stress analyses were also performed to obtain flaw-tip stress distributions for interpretation of the test results. (authors)

Verification tests for GRAD, a computer program to predict nonuniform deformation and failure of Zr-2.5 wt percent Nb pressure tubes during a postulated loss-of-coolant accident

International Nuclear Information System (INIS)

Shewfelt, R.S.W.; Godin, D.P.

1985-03-01

During a postulated loss-of-coolant accident in a CANDU reactor, the temperature of the pressure tubes could rise sufficiently so that ballooning could occur. It is also likely that there would be a variation in temperature around the tube circumference, causing the deformation to be nonuniform. Since the deformation of the pressure tube controls how the core heat is transferred to the surrounding moderator, which is a large heat sink, a computer program, GRAD, has been developed to predict this nonuniform deformation. Numerous biaxial creep tests were done, where the temperature of internally pressurized sections of Zr-2.5 wt percent Nb pressure tubes were ramped to check the ability of GRAD to predict the resulting nonuniform deformation and possible tube failure. GRAD was successful in predicting the average transverse creep strain observed during the tests and the local transverse creep strain at the end of the tests. GRAD was also able to predict the failure time and average transverse creep strain at failure for all the specimens that failed

A model for phase evolution and volume expansion in tube type Nb3Sn conductors

Science.gov (United States)

Xu, X.; Sumption, M. D.; Collings, E. W.

2013-12-01

In this work, an analytic model for phase formation and volume expansion during heat treatment in tube type Nb3Sn strands is presented. Tube type Nb3Sn conductors consist of Nb or Nb-Ta alloy tube with a simple Cu/Sn binary metal insert to form the basic subelement (filament). A number of these elements, each with an outer Cu jacket, are restacked to form a multifilamentary strand. The present tube type conductors, with 4.2 K, 12 T non-Cu critical current density (Jc) in the 2000-2500 A mm-2 range and effective subelement diameters (deff) in the 12-36 μm range, are of interest for a number of applications. During the reaction of typical tube type strands, the Sn-Cu becomes molten and reacts with the Nb tube first to form NbSn2, then Nb6Sn5. At later times in the reaction sequence, all of the NbSn2 and Nb6Sn5 is converted to Nb3Sn. Some of the Nb3Sn is formed by a Nb-Sn reaction and has a fine grain (FG) structure, while some is converted from Nb6Sn5, which results in a coarse grain (CG) region. The fractions of FG and CG A15 are important in determining the final conductor properties. In this work we develop an analytic model to predict the radial extents of the various phases, and in particular the final FG and CG fractions based on the starting Nb, Cu, and Sn amounts in the subelements. The model is then compared to experimental results and seen to give reasonable agreement. By virtue of this model we outline an approach to minimize the CG regions in tube type and PIT strands and maximize the final FG area fractions. Furthermore, the volume change during the various reaction stages was also studied. It is proposed that the Sn content in the Cu-Sn alloy has a crucial influence on the radial expansion.

Influence of hydride orientation on fracture toughness of CWSR Zr-2.5%Nb pressure tube material between RT and 300 °C

Energy Technology Data Exchange (ETDEWEB)

Sharma, Rishi K. [Engineering Directorate, Nuclear Power Corporation of India Limited, Mumbai 400094 (India); Department of Mechanical Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400076 (India); Sunil, Saurav; Kumawat, B.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Singh, R.N., E-mail: rnsingh@barc.gov.in [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Tewari, Asim [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400076 (India); Kashyap, B.P. [Department of Metallurgical Engineering and Materials Science, Indian Institute of Technology Bombay, Powai, Mumbai 400076 (India)

2017-05-15

An experimental setup was designed, fabricated and used to form radial hydrides in Zr-2.5%Nb alloy pressure tube spool. The design of setup was based on ensuring a hoop stress in the spool greater than threshold stress for reorientation of hydrides in this alloy, which was achieved by manipulating the thermal expansion coefficient of the plunger and pressure tube material and diametral interference between them. The experimental setup was loaded on a universal testing machine (UTM) fitted with an environmental chamber and subjected to a temperature cycle for the stress reorientation treatment. The metallographic examination of the hydrogen charged spools subjected to stress re-orientation treatment using this set up revealed formation of predominantly radial hydrides. The variation of fracture toughness of material containing radial hydride with test temperature showed typical ‘S’ curve behavior with transition temperatures more than that of the material containing circumferential hydride.

Predicting crack instability behavior of burst tests from small specimens for irradiated Zr-2.5Nb pressure tubes

International Nuclear Information System (INIS)

Davies, P.H.

1997-01-01

A scaling approach, based on the deformation J-integral at maximum load obtained from small specimens, is proposed for predicting the crack instability behavior of burst tests on irradiated Zr-2.5Nb pressure tubes. An assessment of this approach is carried out by comparison with other toughness criteria such as the modified J-integral and the plastic work dissipation rate approach. The largest discrepancy between the different parameters occurs for materials of intermediate toughness which exhibit the most stable crack growth and tunnelling up to maximum load. A study of one material of intermediate toughness suggests crack-front tunnelling has a significant influence on the results obtained from the 17-mm-wide specimens. It is shown that for a tube of intermediate toughness the different approaches can significantly underpredict the extent of stable crack growth before instability in a burst test even after correcting for tunnelling. The usefulness of a scaling approach in reducing the discrepancy between the small- and large-scale specimen results for this material is demonstrated

Influence of deuterium content on tensile behavior of Zr-2.5Nb pressure tube material in the temperature range of ambient to 300 degC

International Nuclear Information System (INIS)

Bind, A.K.; Singh, R.N.; Chakravartty, J.K.; Dhandharia, Priyesh; Ghosh, Agnish; More, Nitin S.; Chhatre, A.G.; Vijayakumar, S.

2011-08-01

Tensile properties of autoclaved zirconium-2.5 wt. % niobium pressure tube material were evaluated by uniaxial tension tests at temperatures between 25 and 300 degC and under strain-rates of 1.075 x 10 -4 /s. Six number of Zr-2.5Nb alloy pressure tube spools of length 130 mm were obtained from pressure tube number 19-2557-2. Five spools were polished with abrasive paper to remove the oxide layer. These spools were gaseously charged with controlled amount of deuterium. The target deuterium concentrations were 25, 50, 75, 100 and 200 wppm of hydrogen equivalent. Ten samples were machined by EDM wire cutting from every spool. The tensile specimen axis was oriented along longitudinal direction of the tube. Metallographic examination of the deuterium charged samples suggested that the deuterides were predominantly circumferential deuterides. Analysis of tensile results showed that both yield and ultimate tensile strengths of this alloy decreased monotonically with increasing test temperatures. The tensile ductility decreased marginally with increase in test temperature from ambient to 300 degC. It was also observed that both strength and ductility appear to be unaffected by deuterium content at all temperatures, thereby suggesting that at least up to 200 wppm (Heq.) of deuterium tensile properties are not influenced by deuterium. (author)

Predicting the in-reactor mechanical behavior of Zr-2.5Nb pressure tubes from postirradiation microstructural examination data

International Nuclear Information System (INIS)

Griffiths, M.; Davies, P.H.; Davies, W.G.; Sagat, S.

2002-01-01

Postirradiation microstructure examinations of Zr-2.5Nb pressure tubes removed from service in CANDU reactors have shown clear trends in the dislocation structure and the state of the β-phase, as a function of operating temperature, neutron flux, and time. These microstructural parameters correlate well with changes in the mechanical properties. For example, the rapid increase in dislocation loop density in the early stages of irradiation corresponds with a rapid increase in tensile strength and DHC velocity, and a corresponding decrease in fracture toughness. There is also a strong negative correlation between the degree of β-phase decomposition and DHC velocity. In addition to the effects of microstructure evolution on the mechanical properties, changes in the a-type and c-component dislocation loop densities also affect irradiation deformation (creep and growth). Statistical analyses of the irradiation microstructure data have been used to derive empirical relationships between dislocation densities and β-phase structure with temperature, flux, and time. The relationships thus derived are useful in predicting where the mechanical properties are most affected by the in-reactor operating conditions. The predictions are compared with mechanical test data for samples from various axial and circumferential locations of 42 pressure tubes removed from operating CANDU reactors. The results are discussed in terms of the mechanisms controlling tensile strength, fracture, delayed-hydride-cracking, and in-reactor deformation. (author)

Development of Evaluation Technology of the Integrity of HWR Pressure Tubes

Energy Technology Data Exchange (ETDEWEB)

Jeong, Y. M.; Kim, Y. S.; Im, K. S.; Kim, K. S.; Ahn, S. B

2007-06-15

Zr-2.5Nb pressure tubes are one of the most critical structural components governing the lifetime of the heavy water reactors to carry fuel bundles and heavy coolant water inside. Since they are being degraded during their operation in reactors due to dimensional changes caused by creep and irradiation growth, neutron irradiation and delayed hydride cracking, it is required to evaluate their degradation by conducting material testing and examinations on the highly irradiated pressure tubes in hot cells and to keep tracking of their degradation behavior with operation time, which are the aim of this project.

Development and performance of inspection equipment for pressure tubes in Fugen

International Nuclear Information System (INIS)

Naruo, Kazuteru; Tanimoto, Ken-ichi; Ohta, Takeo; Nakamura, Takahisa; Imaizumi, Kiyoshi.

1984-01-01

The pressure tubes of Fugen are the important equipment as the many tubes compose the core, and since they are made of Zr-2.5% Nb alloy which has been used for the first time in Japan, they have become the object of monitoring (the follow-up investigation of the change of inside diameter, the presence of defects and so on) in addition to the in-service inspection. In this paper, on the inspection equipment for pressure tubes, that has been developed independently by the Power Reactor and Nuclear Fuel Development Corp. in order to carry out the ISI and monitoring, the course of development and the construction and the performance are reported, and the results of having used it for the fourth regular inspection of Fugen are described. The 10-year plan of the ISI and monitoring of pressure tubes is shown. The core of Fugen is composed of 224 pressure tubes, therefore, the inspection is carried out by sampling inspection. The monitoring is carried out on four tubes for the follow-up investigation and one tube that shows the severest operation history at the time of inspection. The equipment performs ultrasonic flaw detection, the measurement of inside diameter and the visual inspection of internal surface. (Kako, I.)

Aqueous corrosion behaviour of Zr-1 Nb and Zr-20 Nb with different heat treatments

International Nuclear Information System (INIS)

Jaime Solis, F.; Bordoni, Roberto; Olmedo, Ana M.; Villegas, Marina; Miyagusuku, Marcela

2003-01-01

The corrosion behaviour of Zr-1 Nb and Zr-20 Nb coupons annealed at 850 C degrees during 1 hour and afterwards aged at different temperatures and time periods was studied. The Zr-1 Nb samples were aged at 400 and 500 C degrees and the Zr-20 Nb samples at 265 and 550 C degrees. The results have shown that ageing increases the corrosion resistance because the aged microstructure is somewhat closer to the equilibrium one. This was not the case of Zr-1 Nb aged 72 hs at 400 C degrees. The presence of the ω-phase does not have a deleterious effect in the corrosion behaviour of Zr-20 Nb. Also, an ageing of 2200 h at 265 C degrees induced a relevant decrease in the corrosion rate of Zr-20 Nb indicating a decomposition of the β- Zr phase. This effect was observed at the inlet of pressure tubes in CANDU reactors. The results obtained will be used to establish the relative importance of the α-Zr and β-Zr phases in the corrosion behaviour of pressure tubes. (author)

Probabilistic modeling of material resistance to crack initiation due to hydrided region overloads in CANDU Zr-2.5%Nb pressure tubes

International Nuclear Information System (INIS)

Gutkin, L.; Scarth, D.A.

2014-01-01

Zr-2.5%Nb pressure tubes in CANDU nuclear reactors are susceptible to hydride-assisted cracking at the locations of stress concentration, such as in-service flaws. Probabilistic methodology is being developed to evaluate such flaws for crack initiation due to hydrided region overloads, which occur when the applied stress acting on a flaw with an existing hydrided region at its tip exceeds the stress at which the hydrided region is formed. As part of this development, probabilistic modeling of pressure tube material resistance to overload crack initiation has been performed on the basis of a set of test data specifically designed to study the effects of non-ratcheting hydride formation conditions and load reduction prior to hydride formation. In the modeling framework, the overload resistance is represented as a power-law function of the material resistance to initiation of delayed hydride cracking under constant loading, where both the overload crack initiation coefficient and the overload crack initiation exponent vary with the flaw geometry. In addition, the overload crack initiation coefficient varies with the extent of load reduction prior to hydride formation as well as the number of non-ratcheting hydride formation thermal cycles. (author)

Probabilistic modeling of material resistance to crack initiation due to hydrided region overloads in CANDU Zr-2.5%Nb pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Gutkin, L.; Scarth, D.A. [Kinectrics Inc., Toronto, ON (Canada)

2014-07-01

Zr-2.5%Nb pressure tubes in CANDU nuclear reactors are susceptible to hydride-assisted cracking at the locations of stress concentration, such as in-service flaws. Probabilistic methodology is being developed to evaluate such flaws for crack initiation due to hydrided region overloads, which occur when the applied stress acting on a flaw with an existing hydrided region at its tip exceeds the stress at which the hydrided region is formed. As part of this development, probabilistic modeling of pressure tube material resistance to overload crack initiation has been performed on the basis of a set of test data specifically designed to study the effects of non-ratcheting hydride formation conditions and load reduction prior to hydride formation. In the modeling framework, the overload resistance is represented as a power-law function of the material resistance to initiation of delayed hydride cracking under constant loading, where both the overload crack initiation coefficient and the overload crack initiation exponent vary with the flaw geometry. In addition, the overload crack initiation coefficient varies with the extent of load reduction prior to hydride formation as well as the number of non-ratcheting hydride formation thermal cycles. (author)

Delayed hydride cracking and elastic properties of Excel, a candidate CANDU-SCWR pressure tube material

International Nuclear Information System (INIS)

Pan, Z.L.

2010-01-01

Excel, a Zr alloy which contains 3.5%Sn, 0.8%Nb and 0.8%Mo, shows high strength, good corrosion resistance, excellent creep-resistance and dimension stability and thus is selected as a candidate pressure tube material for CANDU-SCWR. In the present work, the delayed hydride cracking properties (K IH and the DHC growth rates), the hydrogen solubility and elastic modulus were measured in the irradiated and unirradiated Excel pressure tube material. (author)

Crystalline texture study of Zr-2.5%Nb pressure pipes by DRX pole figures

International Nuclear Information System (INIS)

Buioli, C P; Banchik, A D; Vizcaino, P; Samper, R; Testone, S

2012-01-01

This work presents the study of crystalline texture of Zr-2.5%Nb pressure tubes (A, B and C); structural components of CANDU nuclear power reactors [1]. The study of texture was made using the RX diffraction technique, making measurements of direct pole figures [2], in a Phillips diffractometer with pole goniometer (IFIR-CONICET). The texture was determined calculating the coefficients of JJ Kearns [3]. The samples used correspond to transversal sections of the tubes, front and back, and were characterized making de measurements of five pole figures with Miller index (001), (100), (101), (102) and (110), in order to represent in a complete way the texture in the material. The calculated Kearns coefficients were compared with the specifications given by the designer of pressure tubes AECL [4] (author)

A software tool for evaluation of hydrogen ingress in CANDU pressure tubes

International Nuclear Information System (INIS)

Mihalache, Maria; Vasile, Radu; Deaconu, Mariea

2009-01-01

The prediction of hydrogen isotopes concentration into the body and in the rolled joints of operating pressure tubes as a function of reactor hot hours is very important in many fitness-for-service assessments and end of life estimates. The rolled joints are high stress zones with potential for delayed hydride cracking. Predictive models for assessing the long-term deuterium ingress in both body and rolled joint of the pressure tubes have been implemented in a software tool, ROHID, developed in INR-Pitesti. ROHID is a PC-based Windows application with a user-friendly interface that predicts the equivalent hydrogen ingress for Zr-2.5Nb pressure tubes. It uses colour-coded reactor core maps to display the predicted deuterium concentration as a function of time for selected axial locations. Plots of deuterium versus axial location and time for individual pressure tubes are also available. Also, the software tool can predict the exceeding of hydrogen terminal solid solubility (HTSS) from hydrides during precipitation and dissolving processes as a function of time and axial location. (authors)

Pressure tube type reactors

International Nuclear Information System (INIS)

Komada, Masaoki.

1981-01-01

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

High-pressure phases of Weyl semimetals NbP, NbAs, TaP, and TaAs

Science.gov (United States)

Guo, ZhaoPeng; Lu, PengChao; Chen, Tong; Wu, JueFei; Sun, Jian; Xing, DingYu

2018-03-01

In this study, we used the crystal structure search method and first-principles calculations to systematically explore the highpressure phase diagrams of the TaAs family (NbP, NbAs, TaP, and TaAs). Our calculation results show that NbAs and TaAs have similar phase diagrams, the same structural phase transition sequence I41 md→ P6¯ m2→ P21/ c→ Pm3¯ m, and slightly different transition pressures. The phase transition sequence of NbP and TaP differs somewhat from that of NbAs and TaAs, in which new structures emerge, such as the Cmcm structure in NbP and the Pmmn structure in TaP. Interestingly, we found that in the electronic structure of the high-pressure phase P6¯ m2-NbAs, there are coexistingWeyl points and triple degenerate points, similar to those found in high-pressure P6¯ m2-TaAs.

Design and fabrication of Sn-Nb-Cu-Ta-C composites for multifilamentary superconducting Nb/sub 3/Sn wires by using the modified tube technique

Energy Technology Data Exchange (ETDEWEB)

Glowacki, B A; Kosek, Z M

1987-10-01

The factors determining the design and fabrication of Nb/sub 3/Sn multifilamentary wires by the tube technique are discussed. New improved methods of obtaining multifilamentary Nb/sub 3/Sn wires on the basis of both external diffusion and internal diffusion processes, by using the tube technique in a simpler and less expensive way, are presented.

Final report on development evaluation of Task Group 3 pressure tubes

International Nuclear Information System (INIS)

Fleck, R.G.; Price, E.G.; Cheadle, B.A.

1983-11-01

This report describes the production and evaluation of pressure tubes manufactured to the recommendations of Task Group 3 (TG3) of the Creep Engineering Design Plan. The Zr-2.5 wt percent Nb tubes were manufactured by modified production route to change their metallurgical structure and so reduce the in-service elongation rates. Three modified routes were investigated and a total of twenty-eight tubes produced. There were no difficulties in manufacture and the tubes satisfied the quality assurance and design specifications of reactor grade tubes. Metallurgical evaluation showed that the expected changes in microstructure had occurred but not to the extent anticipated. The TG3 tubes were found to have comparable properties to current tubes when tested for: tensile strength (irradiated and unirradiated); hydride cracking; stress to reorient hydrides; hydrogen diffusion; flaw tolerance; corrosion (irradiated and unirradiated); wear; rolled joint characteristics; irradiation creep and growth. Lower in-service elongation rates are expected for tubes produced by two of the modified routes

Pressure tube reactor

International Nuclear Information System (INIS)

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

1993-01-01

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

A review of current knowledge on the effects of hydrogen on the pressure tubes of Ontario Hydro operating reactors

International Nuclear Information System (INIS)

Leger, M.

1982-01-01

Since the occurrence of cracking in Zr-2.5 wt% Nb pressure tubes in Pickering 'A' units 3 and 4 in 1974/75 a great deal of information on the behaviour of hydrogen in pressure tube materials has been generated through research effort by both AECL and Ontario Hydro. In order to use this information effectively and to provide direction and co-ordination for ongoing research, a review of available information and current concerns on hydrogen in pressure tubes was undertaken. The review was divided into two main areas of interest: hydrogen ingress and hydride effects. The uncertainties in the rates of hydrogen ingress into the pressure tubes have been found to be very large. On the basis of current knowledge, predictions of the future behaviour of pressure tubes due to hydride effects are extremely difficult

Pressure tube reactor

International Nuclear Information System (INIS)

Seki, Osamu; Kumasaka, Katsuyuki.

1988-01-01

Purpose: To remove the heat of reactor core using a great amount of moderators at the periphery of the reactor core as coolants. Constitution: Heat of a reactor core is removed by disposing a spontaneous recycling cooling device for cooling moderators in a moderator tank, without using additional power driven equipments. That is, a spontaneous recycling cooling device for cooling the moderators in the moderator tank is disposed. Further, the gap between the inner wall of a pressure tube guide pipe disposed through the vertical direction of a moderator tank and the outer wall of a pressure tube inserted through the guide pipe is made smaller than the rupture distortion caused by the thermal expansion upon overheating of the pressure tube and greater than the minimum gap required for heat shiels between the pressure tube and the pressure tube guide pipe during usual operation. In this way, even if such an accident as can not using a coolant cooling device comprising power driven equipment should occur in the pressure tube type reactor, the rise in the temperature of the reactor core can be retarded to obtain a margin with time. (Kamimura, M.)

Deuterium diffusion along the three principal directions in anisotropic Zr–2.5Nb

Energy Technology Data Exchange (ETDEWEB)

Khatamian, D., E-mail: khatamiand@aecl.ca

2013-12-15

Highlights: •Diffusivity of deuterium in a Zr–2.5Nb plate was measured. •The plate had similar microstructure to that of a CANDU pressure tube. •Diffusion coefficients in the radial direction were 24% less than in the longitudinal direction. •These findings are in close agreement with the theoretical estimates made in the literature. -- Abstract: Pressure tubes of cold-worked Zr–2.5Nb material are used in the CANDU (CANada Deuterium Uranium is a registered trademark of Atomic Energy of Canada Ltd.) reactors to contain the fuel bundles and the heavy water coolant. The pressure tube microstructure consists of α-Zr platelets, with an aspect ratio of 1:5:50 in the radial, transverse and longitudinal directions, surrounded by a thin layer of β-Zr. The diffusivity of hydrogen in β-Zr is higher than in α-Zr. As a result, the diffusivity of hydrogen in the pressure tube material is enhanced compared to unalloyed α-Zr. Due to the anisotropic microstructure of the pressure tube with respect to the β-Zr network, the diffusivity of hydrogen is thought to be different in the three principal pressure tube directions. Measurements made using specimens machined from an 8 mm thick Zr–2.5Nb plate, with a microstructure similar to that of a CANDU pressure tube, suggest the difference between the diffusion constants for deuterium in different directions of an as-installed pressure tube may be as high as 25%.

Effect of heavy ion irradiation and α+β phase heat treatment on oxide of Zr-2.5Nb pressure tube material

Energy Technology Data Exchange (ETDEWEB)

Choudhuri, Gargi, E-mail: gargi@barc.gov.in [Quality Assurance Division, BARC, Mumbai, 400085 (India); Mukherjee, P.; Gayathri, N. [Variable Energy Cyclotron Centre, Kolkata, 700064 (India); Kain, V.; Kiran Kumar, M.; Srivastava, D. [Material Science Division, BARC, Mumbai, 400085 (India); Basu, S. [Solid State Physics Division, BARC, Mumbai, 400085 (India); Mukherjee, D. [Quality Assurance Division, BARC, Mumbai, 400085 (India); Dey, G.K. [Material Science Division, BARC, Mumbai, 400085 (India)

2017-06-15

Effect of heavy-ion irradiation on the crystalline phase transformation of oxide of Zr-2.5Nb alloys has been studied. The steam-autoclaved oxide of pressure tube is irradiated with 306 KeV Ar{sup +9} ions at a dose of 3 × 10{sup 19} Ar{sup +9}/m{sup 2}. The damage profile has been estimated using “Stopping and Range of Ions in Matter” computer program. The variation of the crystal structure along the depth of the irradiated oxide have been characterized non-destructively by Grazing Incidence X-ray Diffraction technique and compared with unirradiated-oxide. The effect of different base metal microstructures on the characteristic of oxide has also been studied. Base metal microstructure as well as the cross-sectional oxide have been characterized using transmission electron microscope. Heavy ion irradiation can significantly alter the distribution of phases in the oxide of the alloy. The difference in chemical state of alloying element has also been found between unirradiated-oxide with that of irradiated-oxide using X-ray photo electron spectroscopy. Chemical state of Nb in steam autoclaved oxide is also altered when the base metal is α + β heat treated.

The creep and stress-rupture behaviour under internal pressure of tubes made from austenitic stainless steel X8 CrNiMoNb 1616 (Material No. 1.4981)

International Nuclear Information System (INIS)

Schaefer, L.; Polifka, F.; Kempe, H.

1979-05-01

Creep and stress rupture tests have been performed at 600, 650, 700 and 750 0 C on tubes made from three different heats from the austenitic stainless steel X8 CrNiMoNb 1616 (Material No. 1.4981). The tubes were loaded by internal pressure and the tangential (hoop) creep strain was measured continuously. The results are presented in form of creep curves, stress-time to rupture curves and curves for a creep limit. The average and minimum creep rates as a function of the applied stress have been evaluated and are described with a creep law analogous to Norton's creep law. An interpolation and extrapolation of the stress-rupture-strength and the creep strength are possible using the time-temperature-parameter-plot after Larson and Miller. (orig.) [de

A technique of taking samples from inside the pressure tube along the axis of the fuel channel

International Nuclear Information System (INIS)

Gyongyosi, T.

2005-01-01

Full text: The ageing process through its complex mechanisms affects in time more or less the component parts, the systems and the structures of the nuclear power plant. For CANDU type nuclear power plant the main component part in operation is the pressure tube, made from Zr - 2,5% Nb alloy, used in extreme hard operation conditions (static and dynamic loading from high pressure and temperature and high neutron flux). The pressure tube endures, in time, changing of the material and of the geometry. Additionally, the excessive hydrogen uptake initiates the Delayed Hydride Cracking (DHC) mechanism on the pressure tubes. For checking the evolution of the hydride / deuteration phenomena in the material of the pressure tube and especially to establish the real lifetime as compared to design lifetime it is useful to initiate, develop and apply a technology and a complex equipment for taking samples directly from inside the pressure tube, this enabling the determination of the hydrogen content. In the paper are showed briefly: - the evolution in time of the techniques for axial taking of the samples from inside of the pressure tube used in the CANDU 6 NPP; - the reasons that determined us to develop one of these technologies; - the technological facilities needed to apply it. (author)

A technique of taking samples from inside the pressure tube along the axis of the fuel channel

International Nuclear Information System (INIS)

Gyongyosi, T.

2005-01-01

The ageing process through its complex mechanisms affects in time more or less the component parts, the systems and the structures of the nuclear power plant. For CANDU type nuclear power plant the main component part in operation is the pressure tube, made from Zr - 2,5% Nb alloy, used in extreme hard operation conditions (static and dynamic loading from high pressure and temperature and high neutron flux). The pressure tube endures, in time, changing of the material and of the geometry. Additionally, the excessive hydrogen uptake initiates the Delayed Hydride Cracking (DHC) mechanism on the pressure tubes. For checking the evolution of the hydride/deuteration phenomena in the material of the pressure tube and especially to establish the real lifetime as compared to design lifetime it is useful to initiate, develop and apply a technology and a complex equipment for taking samples directly from inside the pressure tube, this enabling the determination of the hydrogen content. In the paper are showed briefly: - the evolution in time of the techniques for axial taking of the samples from inside of the pressure tube used in the CANDU 6 NPP; - the reasons that determined us to develop one of these technologies; - the technological facilities needed to apply it. (author)

Study on the manufacturing process, causes of the pressure tube failure and methods for improving its performance

Energy Technology Data Exchange (ETDEWEB)

You, Ho Sik; Jeong, Jin Kon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-07-01

Manufacturing processes of Zr-2.5Nb pressure tube used in CANDU reactor, effects of impurities on the properties of the pressure tube, experiences and causes of the pressure tube cracking accident and the development programs on the fuel channel at AECL have been described. Fabrication processes on the pressure tube have been explained in detail from the sponge production step to the final product. Test methods that are performed to verify the integrity of the final product have also been described. Most of the pressure tube rupture accidents were caused by DHC (Delayed Hydride Cracking). In cases of Pickering units 3 and 4 and Bruce unit 2, excessive residual stresses induced by improper rolled joint process had played a role to cause DHC. In Pickering unit 2, cracks formed by contact between pressure and calandria tubes due to the movement of garter spring were direct cause of failure. After the accidents, a lot of R and D programs on each component of the fuel channel have been carried out. The study on the improvement of manufacturing processes such as increasing cold working rate, performing the intermediate and final annealing and adding the third element like Fe, V, Cr for enhancing the pressure tube performance are on progress. To suppress hydrogen uptake into the pressure tube, the methods such as zirconia coating on the pressure tube, Cr-plating on the end fitting and placing the yttrium getter on the pressure tube are considered. Experiments on each test specimen are currently under way. Owing to such an effort, more advanced fuel channel can be installed in the next CANDU reactor. 6 tabs., 20 figs., 20 refs. (Author).

Creep properties of annealed Zr-Nb-O and stress-relieved Zr-Nb-Sn-Fe cladding tubes and their performance comparison

International Nuclear Information System (INIS)

Ko, S.; Hong, S.I.; Kim, K.T.

2010-01-01

Creep properties of annealed Zr-Nb-O and stress-relieved Zr-Nb-Sn-Fe cladding tubes were studied and compared. The creep rates of the annealed Zr-Nb-O alloy were found to be greater than those of the stress-relieved Zr-Nb-Sn-Fe alloy. Zr-Nb-O alloy was found to have stress exponents of 5-7 independent of stress level whereas Zr-Nb-Sn-Fe alloy exhibited the transition of the stress exponent from 6.5 to 7.5 in the lower stress region to ∼4.2 in the higher stress region. The reduction of stress exponent at high stresses in Zr-Nb-Sn-Fe can be explained in terms of the dynamic solute-dislocation effect caused by Sn atoms. The constancy of stress exponent without the transition was observed in Zr-Nb-O alloy, supporting that the decrease of the stress exponent with increasing stress in Zr-Nb-Sn-Fe is associated with Sn atoms. The difference of creep life between annealed Zr-Nb-O and stress-relieved Zr-Nb-Sn-Fe is not large considering the large difference of strength level between annealed Zr-Nb-O and annealed stress-relieved Zr-Nb-Sn-Fe. The better-than-expected creep life of annealed Zr-Nb-O alloy can be attributable to the combined effects of creep ductility enhancement associated with softening and the decreased contribution of grain boundary diffusion due to the increased grain size.

The elastic properties of zirconium alloy fuel cladding and pressure tubing materials

International Nuclear Information System (INIS)

Rosinger, H.E.; Northwood, D.O.

1979-01-01

A knowledge of the elastic properties of zirconium alloys is required in the mathematical modelling of cladding and pressure tubing performance. Until recently, little of this type of data was available, particularly at elevated temperatures. The dynamic elastic moduli of zircaloy-2, zircaloy-4, the alloys Zr-1.0 wt%Nb, Zr-2.5 wt%Nb and Marz grade zirconium have therefore been determined over the temperature range 275 to 1000 K. Young's modulus and shear modulus for all the zirconium alloys decrease with temperature and are expressed by empirical relations fitted to the data. The elastic properties are texture dependent and a detailed study has been conducted on the effect of texture on the elastic properties of Zr-1.0 wt% Nb over the temperature range 275 to 775 K. The results are compared with polycrystalline elastic constants computed from single crystal elastic constants, and the effect of texture on the dynamic elastic moduli is discussed in detail. (Auth.)

Evaluation of a leaking crack in an irradiated CANDU pressure tube

International Nuclear Information System (INIS)

Coleman, C.E.; Simpson, L.A.

1988-06-01

Leak-before-break is used in CANDU reactors as part of the defence against rupture of the pressure tubes. Two important features of this technique are the action time available for detection of a leaking crack and the size of the leak allowing crack location. Support for continued reliance on leak-before-break is being obtained from experiments, on irradiated Zr-2.5 Nb pressure tubes attached to their end fittings, that simulate the behaviour of a leaking crack in a reactor. At reactor operating temperatures leaking cracks grow more slowly than dry cracks in the laboratory because they are cooled when pressurised water flashes to steam on their surface. These cracks remain stable till they are at least 70 mm long. From the results of these experiments the action time is at least 100 h. The leak rate increases rapidly when a through-wall crack extends a small amount, thus greatly assisting with crack location

Pressure tube type research reactor

International Nuclear Information System (INIS)

Ueda, Hiroshi.

1976-01-01

Object: To prevent excessive heat generation due to radiation of a pressure tube vessel. Structure: A pressure tube encasing therein a core comprises a dual construction comprising inner and outer tubes coaxially disposed. High speed cooling water is passed through the inner tube for cooling. In addition, in the outer periphery of said outer tube there is provided a forced cooling tube disposed coaxially thereto, into which cooling fluid, for example, such as moderator or reflector is forcibly passed. This forced cooling tube has its outer periphery surrounded by the vessel into which moderator or reflector is fed. By the provision of the dual construction of the pressure tube and the forced cooling tube, the vessel may be prevented from heat generation. (Ikeda, J.)

Creep and stress rupture behaviour of zircaloy-2 and Zr-2.5% Nb alloy tubes at 573 K

International Nuclear Information System (INIS)

Laha, K.; Bhanu Sankara Rao, K.; Chandravathi, K.S.; Mannan, S.L.

1992-01-01

Zirconium alloys are extensively used for coolant tubes of pressurised heavy water reactors. The choice of these materials is based on their good corrosion resistance in water, low capture cross section for thermal neutrons and good mechanical properties. In this paper the results of an investigation performed on the creep and rupture behaviour of indigenously produced zircaloy-2 and Zr-2.5% Nb alloy are presented. Samples for creep testing were cut longitudinally from finished pressure tubes. Creep rupture tests were carried out in air under constant load conditions at 300 C employing five stress levels in the range 300-360 MPa. Zr-2.5% Nb alloy displayed higher rupture lives at all stress levels compared to zircaloy-2. Steady state creep rate of Zr-2.5%Nb was lower than that zircaloy-2 at identical stress levels. In the stress range of the experiments, the dependence of the steady state creep rate (ε s ) on applied stress (σ) for both the alloys could be represented by a power law, ε s =A σ n The stress sensitivity (n) for Zr-2.5% Nb was lower than that of zircaloy-2. For both the alloys the time to creep rupture t r was found related to the steady state creep rate through the modified Monkman-Grant relation (ε s ) α . t r = constant. Similar value of α was obtained for both the materials. Zr-2.5%Nb exhibited higher ductility (% elongation to rupture) compared to zircaloy-2 at stress levels ≥ 320 MPa. At lower stresses significant difference in ductility was not noticed. Percentage reduction in area was lower in Zr-2.5%Nb at all stress levels indicating better resistance for necking. The time for onset of tertiary was longer for Zr-2.5% Nb alloy. The proportion of life spent by Zr-2.5% Nb in steady state creep regime was higher compared to that of zircaloy-2. Metallographic investigations on longitudinal sections in both the alloys showed large number of intragranular pores close to the fracture surface. A few number of cracks which are characteristic of

Experimental determination of thermal contact conductance between pressure and calandria tubes of Indian pressurised heavy water reactors

Energy Technology Data Exchange (ETDEWEB)

Dureja, A.K., E-mail: akdureja@barc.gov.in [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Pawaskar, D.N.; Seshu, P. [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Mumbai (India); Sinha, S.K. [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Sinha, R.K. [Department of Atomic Energy, OYC, Near Gateway of India, Mumbai (India)

2015-04-01

Highlights: • We established an experimental facility to measure thermal contact conductance between disc shaped specimens. • We measured thermal contact conductance between Zr-2.5Nb alloy pressure tube (PT) material and Zr-4 calandria tube (CT) material. • We concluded that thermal contact conductance is a linear function of contact pressure for interface of PT and CT up to 10 MPa contact pressure. • We concluded that thermal contact conductance is a weak function of interface temperature. - Abstract: Thermal contact conductance (TCC) is one of the most important parameters in determining the temperature distribution in contacting structures. Thermal contact conductance between the contacting structures depends on the mechanical properties of underlying materials, thermo-physical properties of the interstitial fluid and surface condition of the structures coming in contact. During a postulated accident scenario of loss of coolant with coincident loss of emergency core cooling system in a tube type heavy water nuclear reactor, the pressure tube is expected to sag/balloon and come in contact with outer cooler calandria tube to dissipate away the heat generated to the moderator. The amount of heat thus transferred is a function of thermal contact conductance and the nature of contact between the two tubes. An experimental facility was designed, fabricated and commissioned to measure thermal contact conductance between pressure tube and calandria tube specimens. Experiments were conducted on disc shaped specimens under axial contact pressure in between mandrels. Experimental results of TCC and a linear correlation as a function of contact pressure have been reported in this paper.

Niobium alloys production with elements of high steam pressure and high ductilidate Nb46,5%Ti, Nb 1%Zr, Nb 1%Ti and Nb20% Ta

International Nuclear Information System (INIS)

Pinatti, D.G.; Baldan, C.A.; Dainesi, C.R.; Sandim, H.R.Z.

1988-01-01

The melting technology of niobium alloys with high ductilidade and high steam pressure, having the Ti, Zr and Ta as alloying elements is described. The electron beam technique for production of Nb 46,5%Ti, Nb 1%Zr and Nb 20%Ta alloys is analysed, aiming a product with high grade and low cost. (C.G.C.) [pt

Mechanical properties of Zr-2.5Nb pressure tube material manufactured employing forging routes for PHWR700 - Part 1: tensile behavior

International Nuclear Information System (INIS)

Bind, A.K.; Singh, R.N.; Sunil, Saurav; Chakravartty, J.K.; Ghosh, Agnish; Dhandharia, Priyesh; More, Nitin S.; Vijayakumar, S.; Chhatre, A.G.

2012-01-01

Tensile properties of Zr-2.5Nb alloy tubes produced employing forging to break the cast structure were evaluated by carrying out uni-axial tension test at temperatures between 25 and 325 degC and under strain-rate of 1.075 x 10 -4 /s for both longitudinal and transverse specimens. Both strength and elongation values were comparable for the samples obtained from front and back end of the tube. Transverse samples showed higher strength and lower uniform elongation values as compared to longitudinal samples. The yield strength of double forged material at 25 degC is higher than the PHWR700 specification of a maximum value of 586 MPa. (author)

PPFAE preparation for pressure pipes production to CNE

International Nuclear Information System (INIS)

Azzinari, D; Carus, G; Sabio Calvete, M; Ilarri, M; Segato, M; Perazzo, M; Duarte; J; Duarte, A; Rojas, J; Fleitas, G; Szieber, W C; Bianchi, A D; Flores, A V; Bianchi, D R; Vizcaino, P

2012-01-01

The Life Extension Program for the PHWR Embalse power reactor of Argentina includes the replacement of de original set of 2,5% Nb pressure tubes by a new one It was also decided to made the new set de pressure tubes locally from imported Zr-2,5% Nb extruded tubes. Based on the degraded processes that develop in service a new set of metallurgical requirements was set by the Designer to be follow for new suppliers. As the mechanical properties of pressure tubes in operation depend on the micro-structural deformation, it is necessary to add to the normal macroscopic quality control same additional mechanical and microstructural studies, especially during the development stage. In the present paper the current status of delay hydrogen cracking, full width at half maximum (FWHM) of x-ray diffraction peaks, alpha phase thickness log-normal distributions and Kearns texture factors are summarized to estimate the magnitude of the micro-structural characteristics of pressure tubes obtained from imported Zr-2,5% Nb extruded tubes that were processes following designer specifications. The pressure tubes were obtained by two step of plastic deformation in a HPTR 60-120 tube rolling machine without intermediate annealing and a final thermal treatment (author)

The influence of post-extrusion thermomechanical treatments on the tensile properties of Zr-2.5 wt% Nb alloy

International Nuclear Information System (INIS)

Fleck, R.G.; Shek, G.K.

1983-01-01

The production routes used for Zr-2.5 wt% Nb pressure tubes are described. Tensile results (UTS) from laboratory tests which simulated modified production routes are presented and compared to UTS values of actual pressure tubes. Strengthening of stress relieved Zr-2.5 wt% Nb is discussed in terms of: a) sub-grain formation; b) transformation of the second phase; and c) reorientation of the second phase relative to the matrix. The strength of cold worked Zr-2.5 wt% Nb is not influenced by the prior cold worked grain size. (author)

Effect of texture on creep deformation behavior of Zr-2.5Nb alloy

International Nuclear Information System (INIS)

Guguloth, Krishna; Swaminathan, J.; Mitra, Rahul; Ghosh, R.N.; Singh, R.N.; Chakravartty, J.K.

2016-01-01

Zr-2.5%Nb alloys are extensively used as high temperature pressure tubes in nuclear reactor. Therefore creep behavior of this alloy is of considerable importance. The paper presents creep strain-time plots on two sets of specimens made from two as received pressure tubes having different diameters. These tubes were reported to have undergone different processing routes; both tubes were autoclaved at the same temperature in the steam atmosphere. A comparison of the creep strain-time plots of the two sets of specimen under identical test conditions showed a marked difference. The chemical composition and the microstructure of the two sets of samples were also found to be similar. Therefore X-ray diffraction patterns were taken from the two tubes. The ratio of intensity of two prominent reflections from 0002 and 1120 planes of alpha Zr in the case of 90mm tube was found to be 1.79; whereas that from the 110mm tube was 0.25. This suggests that in the case of 110mm tube most of the basal planes were less favorably oriented with respect to the loading axis. This is the reason why creep strength of 110mm tube was found to be higher. The paper also describes how the effect of texture can be incorporated in evaluating the creep behavior of Zr-Nb alloy. This suggests that a relatively larger volume of creep test data on Zr-2.5Nb pressure tube is necessary to account for the effect texture so that a reliable estimate of its creep life could be obtained. (author)

Some engineering aspects of the investigation into the cracking of pressure tubes in the Pickering reactors

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Towgood, G.R.; Hunter, T.A.

1976-01-01

In August 1974, Pickering Unit 3 (514 MWe) was shutdown for a period of 8 months because of cracks in 17 of the 390 pressure tubes. The cracks were a result of incorrect installation procedures during construction. Improper positioning of the rolling tool used to join the Zr-2.5 wt% Nb pressure tube to the end fitting produced very high residual tensile stresses. High stresses in combination with periods with the tubes cold caused the cracking. Crack propagation was by fracture of hydrides which are brittle when cold. Subsequent investigation confirmed that properly rolled joints are not susceptible to such cracking. The resources of Canadian industry, Ontario Hydro and Atomic Energy of Canada were coordinated to find engineering solutions to the crack program. The defective tubes were removed from reactor, thoroughly examined to identify the cause of the cracks, and thoroughly tested to prove safety. Non-destructive techniques were quickly adopted for inspection of tubes in Pickering. Tools and procedures for retubing the 17 channels were prepared and Pickering Unit 3 was returned to service at the end of March 1975. (author)

Quantitative texture determination in pressure tube (Zr-2.5 Wt% Nb alloy) material as a function of cold work

International Nuclear Information System (INIS)

Dey, G.K.; Tewari, R.; Srivastava, D.; De, P.K.; Banerjee, S.; Kiran Kumar, M.; Samajdar, I.

2003-06-01

The texture studies on the pressure tube Zr-2.5 Nb alloy have mainly been confined to the determination of the basal pole distribution along certain direction or the inverse pole presentation in the material. This information though useful does not provide an insight into micro-textural development upon cold working. In the present study, complete bulk as well as micro texture development as a function of cold work has been obtained by determining orientation distribution function. In this work, two distinct starting microstructures of Zr-2.5 wt% Nb have been used -(a) single-phase α(hcp) martensitic structure and (b) two-phase, β(bcc) + α, Widmanstaetten structure. In the second case, the α phase was present in lamellar morphology and β stringers were sandwiched between these a lamella. In some instances single-phase α were present. However, both microstructures had similar starting crystallographic texture. Samples were deformed by unidirectional and cross rolling at room temperature. In the two-phase structure the changes in the bulk texture on cold rolling was found to be insignificant, while in the single-phase material noticeable textural changes were observed. Taylor type deformation texture models predicted textural changes in single-phase structure but failed to predict the observed lack of textural development in the two-phase material. Microtexture observations showed that a plates remained approximately single crystalline after cold rolling, while the β matrix underwent significant orientational changes. Based on microstructural and microtextural observations, a simple model is proposed in which the plastic flow is mainly confined to the β matrix within which the α plates are subjected to in-plane rigid body rotation. The model explains the observed lack of textural developments in the two-phase structure. (author)

Powder-in-Tube (PIT) Nb3Sn conductors for high-field magnets

NARCIS (Netherlands)

Lindenhovius, J.H.; Hornsveld, E.M.; den Ouden, A.; Wessel, Wilhelm A.J.; ten Kate, Herman H.J.

2000-01-01

New Nb3Sn conductors, based on the powder-in-tube (PIT) process, have been developed for application in accelerator magnets and high-field solenoids. For application in accelerator magnets, SMI has developed a binary 504 filament PIT conductor by optimizing the manufacturing process and adjustment

Method of repairing pressure tube type reactors

International Nuclear Information System (INIS)

Asada, Takashi.

1983-01-01

Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

Contrastive Analysis and Research on Negative Pressure Beam Tube System and Positive Pressure Beam Tube System for Mine Use

Science.gov (United States)

Wang, Xinyi; Shen, Jialong; Liu, Xinbo

2018-01-01

Against the technical defects of universally applicable beam tube monitoring system at present, such as air suction in the beam tube, line clogging, long sampling time, etc., the paper analyzes the current situation of the spontaneous combustion fire disaster forecast of mine in our country and these defects one by one. On this basis, the paper proposes a research thought that improving the positive pressure beam tube so as to substitute the negative pressure beam tube. Then, the paper introduces the beam tube monitoring system based on positive pressure technology through theoretical analysis and experiment. In the comparison with negative pressure beam tube, the paper concludes the advantage of the new system and draws the conclusion that the positive pressure beam tube is superior to the negative pressure beam tube system both in test result and test time. At last, the paper proposes prospect of the beam tube monitoring system based on positive pressure technology.

Hydrogen concentration determination in pressure tube samples using differential scanning calorimetry (dsc)

International Nuclear Information System (INIS)

Marinescu, R.; Mincu, M.

2015-01-01

Zirconium alloys are widely used as a structural material in nuclear reactors. It is known that zirconium based cladding alloys absorb hydrogen as a result of service in a pressurized water reactor. Hydrogen absorbed (during operation of the reactor) in the zirconium alloy, out of which the pressure tube is made, is one of the major factors determining the life time of the pressure tube. For monitoring the hydrides, samples of the pressure tube are periodically taken and analyzed. At normal reactor operating temperature, hydrogen has limited solubility in the zirconium lattice and precipitates out of solid solution as zirconium hydride when the solid solubility is exceeded. As a consequences material characterization of Zr-2.5Nb CANDU pressure tubes is required after manufacturing but also during the operation to assess its structural integrity and to predict its behavior until the next in-service inspection. Hydrogen and deuterium concentration determination is one of the most important parameters to be evaluated during the experimental tests. Hydrogen present in zirconium alloys has a strong effect of weakening. Following the zirconium-hydrogen reaction, the resulting zirconium hydride precipitates in the mass of material. Weakening of the material, due to the presence of 10 ppm of precipitated hydrogen significantly affects some of its properties. The concentration of hydrogen in a sample can be determined by several methods, one of them being the differential scanning calorimetry (DSC). The principle of the method consists in measuring the difference between the amount of heat required to raise the temperature of a sample and a reference to a certain value. The experiments were made using a TA Instruments DSC Q2000 calorimeter. This paper contains experimental work for hydrogen concentration determination by Differential Scanning Calorimetry (DSC) method. Also, the reproducibility and accuracy of the method used at INR Pitesti are presented. (authors)

An intermetallic powder-in-tube approach to increased flux-pinning in Nb3Sn by internal oxidation of Zr

Science.gov (United States)

Motowidlo, L. R.; Lee, P. J.; Tarantini, C.; Balachandran, S.; Ghosh, A. K.; Larbalestier, D. C.

2018-01-01

We report on the development of multifilamentary Nb3Sn superconductors by a versatile powder-in-tube technique (PIT) that demonstrates a simple pathway to a strand with a higher density of flux-pinning sites that has the potential to increase critical current density beyond present levels. The approach uses internal oxidation of Zr-alloyed Nb tubes to produce Zr oxide particles within the Nb3Sn layer that act as a dispersion of artificial pinning centres (APCs). In this design, SnO2 powder is mixed with Cu5Sn4 powder within the PIT core that supplies the Sn for the A15 reaction with Nb1Zr filament tubes. Initial results show an average grain size of ˜38 nm in the A15 layer, compared to the 90-130 nm of typical APC-free high-J c strands made by conventional PIT or Internal Sn processing. There is a shift in the peak of the pinning force curve from H/H irr of ˜0.2 to ˜0.3 and the pinning force curves can be deconvoluted into grain boundary and point-pinning components, the point-pinning contribution dominating for the APC Nb-1wt%Zr strands.

Structural and Transport Properties of the Weyl Semimetal NbAs at High Pressure

International Nuclear Information System (INIS)

Zhang Jun; Liu Feng-Liang; Dong Jin-Kui; Xu Yang; Li Shi-Yan; Li Na-Na; Yang Wen-Ge

2015-01-01

We perform a series of high-pressure synchrotron x-ray diffraction (XRD) and resistance measurements on the Weyl semimetal NbAs. The crystal structure remains stable up to 26 GPa according to the powder XRD data. The resistance of NbAs single crystal increases monotonically with pressure at low temperature. Up to 20 GPa, no superconducting transition is observed down to 0.3 K. These results show that the Weyl semimetal phase is robust in NbAs, and applying pressure may not be a good way to obtain a topological superconductor from Weyl semimetal NbAs. (paper)

Stress-strain effects in alumina-Cu reinforced Nb3Sn wires fabricated by the tube process

International Nuclear Information System (INIS)

Murase, Satoru; Nakayama, Shigeo; Masegi, Tamaki; Koyanagi, Kei; Nomura, Shunji; Shiga, Noriyuki; Kobayashi, Norio; Watanabe, Kazuo.

1997-01-01

In order to fabricate a large-bore, high-field magnet which achieves a low coil weight and volume, a high strength compound superconducting wire is required. For those demands we have developed the reinforced Nb 3 Sn wire using alumina dispersion strengthened copper (alumina-Cu) as a reinforcement material and the tube process of the Nb 3 Sn wire fabrication. The ductility study of the composites which consisted of the reinforcement, Nb tube, Cu, and Cu clad Sn brought a 1 km long alumina-Cu reinforced Nb 3 Sn wire successfully. Using fabricated wires measurements and evaluations of critical current density as parameters of magnetic field, tensile stress, tensile strain, and transverse compressive stress, and those of stress-strain curves at 4.2 K were performed. They showed superior performance such as high 0.3% proof stress (240 MPa at 0.3% strain) and high maximum tolerance stress (320 MPa) which were two times as large as those of conventional Cu matrix Nb 3 Sn wire. The strain sensitivity parameters were obtained for the reinforced Nb 3 Sn wire and the Cu matrix one using the scaling law. Residual stress of the component materials caused by cooling down to 4.2 K from heat-treatment temperature was calculated using equivalent Young's modulus, equivalent yield strength, thermal expansion coefficient and other mechanical parameters. Calculated stress-strain curves at 4.2 K for the reinforced Nb 3 Sn wire and the Cu matrix one based on calculation of residual stress, had good agreement with the experimental values. (author)

Fast fracture of a zirconium alloy pressure tube: cause and implications

International Nuclear Information System (INIS)

Price, E.G.; Cheadle, B.A.

1985-12-01

The cause of the unstable fracture of a Zircaloy-2 pressure tube in the core of a CANDU reactor is reviewed. Failure was associated with the presence of brittle zones of zirconium hydride which developed as a result of thermal gradient induced hydrogen diffusion. Unstable fracture occurred when the partial thickness crack reached an unstable length and the crack ran 2 meters along the tube and terminated by circumferential tearing. The partial thickness defect initiated and propagated to an unstable length by delayed hydride cracking is high compared to fatigue progression and increases exponentially with temperature. Delayed hydride cracking can be prevented by reducing residual stresses to a minimum and by high standards of non-destructive testing that ensures freedom from unacceptable defects. Future prevention of fast fracture is based upon the inspection of a limited number of fuel channels for the presence of defects and for conditions which can cause hydride build-up together with the periodic removal of Zr-2.5wt% Nb tubes to monitor their condition

Superconductivity at high pressure in NbSe3

International Nuclear Information System (INIS)

Nunez Regueiro, M.; Castello, D.; Mignot, J.M.

1992-01-01

We have measured the electrical resistivity of NbSe 3 between 2 K and room temperature up to a pressure of 7.2 GPa. At P 1 = 3.5 GPa we observe the extinction of the high-temperature charge density wave (T 1 -CDW) and the enhancement of the superconducting critical temperature T c to ≅ 5 K. The logarithmic pressure slopes of T 1 (P 1 ) and T c (P > P 1 ) are found to be practically equal. A similar behaviour had been reported previously at lower pressures for T 2 (P 2 ) and T c (P 2 1 ) in the distorted state. We discuss these results in terms of an anisotropic superconducting state in NbSe 3 , with different gaps associated with different types of chains. 10 refs., 2 figs

N Reactor pressure tube 1350 postirradiation examination

International Nuclear Information System (INIS)

Cook, D.J.

1977-01-01

The N Reactor pressure tubes were fabricated from Zircaloy-2 primarily due to the excellent corrosion resistance, low neutron absorption, and high strength properties of this alloy. Irradiation damage mechanisms increase the strength and decrease the ductility of the Zircaloy-2. Irradiation data available at the time the tubes were installed indicated that fast neutron irradiation damage mechanisms would not decrease the ductility to unacceptable levels over the estimated plant life of 25 to 30 years. However, because the tubes are a primary coolant system component and only limited data are available on irradiation effects at high fluences, a Postirradiation Examination (PIE) program was developed to assure that service factors do not compromise pressure tube integrity essential to reactor safety. The PIE program requires that a pressure tube be periodically removed from the reactor for destructive testing. The N Reactor Technical Specifications specify that the frequency of pressure tube removal and examination be based upon the previous PIE test results. Four pressure tubes were examined before tube 1350, and the test results were summarized in individual reports. PIE results on tube 1350 were summarized along with the test results on the previous four tubes in a previous report. The purpose of this report is to present in detail the results on PIE of pressure tube 1350, and, in particular, document the technique by which the fracture toughness of the pressure tube was determined

Dynamics of explosively imploded pressurized tubes

Science.gov (United States)

Szirti, Daniel; Loiseau, Jason; Higgins, Andrew; Tanguay, Vincent

2011-04-01

The detonation of an explosive layer surrounding a pressurized thin-walled tube causes the formation of a virtual piston that drives a precursor shock wave ahead of the detonation, generating very high temperatures and pressures in the gas contained within the tube. Such a device can be used as the driver for a high energy density shock tube or hypervelocity gas gun. The dynamics of the precursor shock wave were investigated for different tube sizes and initial fill pressures. Shock velocity and standoff distance were found to decrease with increasing fill pressure, mainly due to radial expansion of the tube. Adding a tamper can reduce this effect, but may increase jetting. A simple analytical model based on acoustic wave interactions was developed to calculate pump tube expansion and the resulting effect on the shock velocity and standoff distance. Results from this model agree quite well with experimental data.

Pressure loss characteristics of LSTF steam generator heat-transfer tubes. Pressure loss increase due to tube internal instruments

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro

1994-11-01

The steam generator of the Large-Scale Test Facility (LSTF) includes 141 heat-transfer U-tubes with different lengths. Six U-tubes among them are furnished with 15 or 17 probe-type instruments (conduction probe with a thermocouple; CPT) protuberant into the primary side of the U-tubes. Other 135 U-tubes are not instrumented. This results in different hydraulic conditions between the instrumented and non-instrumented U-tubes with the same length. A series of pressure loss characteristics tests was conducted at a test apparatus simulating both types of U-tube. The following pressure loss coefficient (K CPT ) was reduced as a function of Reynolds number (Re) from these tests under single-phase water flow conditions. K CPT =0.16 5600≤Re≤52820, K CPT =60.66xRe -0.688 2420≤Re≤5600, K CPT =2.664x10 6 Re -2.06 1371≤Re≤2420. The maximum uncertainty is 22%. By using these results, the total pressure loss coefficients of full length U-tubes were estimated. It is clarified that the total pressure loss of the shortest instrumented U-tube is equivalent to that of the middle-length non-instrumented U-tube and also that a middle-length instrumented U-tube is equivalent to the longest non-instrumented U-tube. Concludingly. it is important to take account of the CPT pressure loss mentioned above in estimation of fluid behavior at the non-instrumented U-tubes either by using the LSTF experiment data from the CPT-installed U-tubes or by using any analytical codes. (author)

Nb effect on Zr-alloy oxidation under high pressure steam at high temperatures

International Nuclear Information System (INIS)

Park, Kwangheon; Yang, Sungwoo; Kim, Kyutae

2005-01-01

The high-pressure steam effects on the oxidation of Zircaloy-4 (Zry-4) and Zirlo (Zry-1%Nb) claddings at high temperature have been analyzed. Test temperature range was 700-900degC, and pressures were 1-150 bars. High pressure-steam enhances oxidation of Zry-4, and the dependency of enhancement looks exponential to steam pressure. The origin of the oxidation enhancement turned out to be the formation of cracks in oxide. The loss of tetragonal phase by high-pressure steam seems related to the crack formation. Addition of Nb as an alloying element to Zr alloy reduces significantly the steam pressure effects on oxidation. The higher compressive stresses and the smaller fraction of tetragonal oxides in Zry-1%Nb seem to be the diminished effect of high-pressure steam on oxidation. (author)

Selected reading on introduction to pressure tube technology

International Nuclear Information System (INIS)

Causey, A.R.; Coleman, C.E.; Ells, C.E.

1981-10-01

Four lectures on pressure tube technology were presented at Sheridan Park, Ontario, on 1981 June 1. The titles were 'Pressure Tubes and Their Operational Environment', 'Fabrication, Inspection and Properties of Current Production Pressure Tubes', 'In-Reactor Deformation of Fuel Channels', and 'Potential Failure Modes in Pressure Tubes'. This report lists the references used in preparing the lectures. It is intended to provide a starting point in reading for people who need to become familiar with pressure tube technology but have little prior knowledge of the topic

Effect of Crack Tip Stresses on Delayed Hydride Cracking in Zr-2.5Nb Tubes

International Nuclear Information System (INIS)

Kim, Young Suk; Cheong, Yong Moo

2007-01-01

Delayed hydride cracking (DHC) tests have shown that the DHC velocity becomes faster in zirconium alloys with a higher yield stress. To account for this yield stress effect on the DHC velocity, they suggested a simple hypothesis that increased crack tip stresses due to a higher yield stress would raise the difference in hydrogen concentration between the crack tip and the bulk region and accordingly the DHC velocity. This hypothesis is also applied to account for a big leap in the DHC velocity of zirconium alloys after neutron irradiation. It should be noted that this is based on the old DHC models that the driving force for DHC is the stress gradient. Puls predicted that an increase in the yield stress of a cold worked Zr-2.5Nb tube due to neutron irradiation by about 300 MPa causes an increase of its DHC velocity by an order of magnitude or 2 to 3 times depending on the accommodation energy values. Recently, we proposed a new DHC model that a driving force for DHC is not the stress gradient but the concentration gradient arising from the stress-induced precipitation of hydrides at the crack tip. Our new DHC model and the supporting experimental results have demonstrated that the DHC velocity is governed primarily by hydrogen diffusion at below 300 .deg. C. Since hydrogen diffusion in Zr-2.5Nb tubes is dictated primarily by the distribution of the β-phase, the DHC velocity of the irradiated Zr-2.5Nb tube must be determined mainly by the distribution of the β-phase, not by the increased yield stress, which is in contrast with the hypothesis of the previous DHC models. In short, a controversy exists as to the effect on the DHC velocity of zirconium alloys of a change in the crack tip stresses by irradiation hardening or cold working or annealing. The aim of this study is to resolve this controversy and furthermore to prove the validity of our DHC model. To this end, we cited Pan et al.'s experiment where the delayed hydride cracking velocity, the tensile strengths

Size determinations, by ultrasonic techniques, of cracks in hydride blisters formed in Zr-2.5 % Nb pressure tubes

International Nuclear Information System (INIS)

Trujillo Badillo, Giovanna; Desimone, Carlos; Domizzi, Gladys

1999-01-01

Non destructive techniques (NDT) are very useful in the detection of flaws produced in structural components in service. During the service of CANDU nuclear power reactors, it is possible that pressure tubes (PT) may contact calandria tubes (CT). After the PT/CT contact, zirconium hydride blisters may form at the point of contact depending on the concentration of hydrogen/deuterium. Zirconium hydride is brittle and is therefore prone to cracking under stress. Ultrasonic NDT is routinely use during PT in service inspection. In order to be able of detecting cracked blisters, it is of great importance the development of standards to calibrate the employed equipment. On this purpose, hydride blisters were grown, in laboratory, on sections of pressure tube. The cracks in the blisters were detected and measured by ultrasonic techniques. The obtained results were compared with measurements carried out in optic microscope, on successive sections of the samples. The crack tip diffraction technique was found to be the more effective for the mentioned ends. (author)

Fabrication of fine-grain tantalum diffusion barrier tube for Nb3Sn conductors

International Nuclear Information System (INIS)

Hartwig, K. T.; Balachandran, S.; Mezyenski, R.; Seymour, N.; Robinson, J.; Barber, R. E.

2014-01-01

Diffusion barriers used in Nb 3 Sn wire are often fabricated by wrapping Ta sheet into a tube with an overlap seam. A common result of such practice is non-uniform deformation in the Ta sheet as it thins by wire drawing because of non-uniform grain size and texture in the original Ta sheet. Seamless Ta tube with a fine-grain and uniform microstructure would be much better for the diffusion barrier application, but such material is expensive and difficult to manufacture. This report presents results on a new fabrication strategy for Ta tube that shows promise for manufacture of less costly tube with an improved microstructure. The fabrication method begins with seam-welded tube but gives a fine-grain uniform microstructure with little difference between the longitudinal seam weld region and the parent metal after post-weld processing. Severe plastic deformation processing (SPD) applied by area reduction extrusion and tube equal channel angular extrusion (tECAE) are used to refine and homogenize the microstructure. Microstructure and mechanical property results are presented for Ta tubes fabricated by this new processing strategy

Fabrication of fine-grain tantalum diffusion barrier tube for Nb3Sn conductors

Science.gov (United States)

Hartwig, K. T.; Balachandran, S.; Mezyenski, R.; Seymour, N.; Robinson, J.; Barber, R. E.

2014-01-01

Diffusion barriers used in Nb3Sn wire are often fabricated by wrapping Ta sheet into a tube with an overlap seam. A common result of such practice is non-uniform deformation in the Ta sheet as it thins by wire drawing because of non-uniform grain size and texture in the original Ta sheet. Seamless Ta tube with a fine-grain and uniform microstructure would be much better for the diffusion barrier application, but such material is expensive and difficult to manufacture. This report presents results on a new fabrication strategy for Ta tube that shows promise for manufacture of less costly tube with an improved microstructure. The fabrication method begins with seam-welded tube but gives a fine-grain uniform microstructure with little difference between the longitudinal seam weld region and the parent metal after post-weld processing. Severe plastic deformation processing (SPD) applied by area reduction extrusion and tube equal channel angular extrusion (tECAE) are used to refine and homogenize the microstructure. Microstructure and mechanical property results are presented for Ta tubes fabricated by this new processing strategy.

Preliminary study for the microstructural characterization of Zr-2.5Nb pressure pipes by electron microscopy

International Nuclear Information System (INIS)

Saumell M, Lani; Oroza Z E, Celiz; Bozzano, P.B; Versaci, R.A

2012-01-01

The Embalse Nuclear Power Station (CNE) is a CANadian Deuterium Uranium (CANDU) Reactor. One of the acceptance criteria for the Zr-2,5Nb alloy pressure tubes of the CNE is based on the width of the α-Zr phase. This width is manually measured, using image processing software. The first part of this paper is oriented to weigh the human factor of these measures and find a proper image treatment to minimize this error. In the second part an automatic procedure is proposed, in order to measure the width of the α-Zr phase and become independent on the human factor (author)

N Reactor pressure tube 2566 postirradiation examination

International Nuclear Information System (INIS)

Scott, K.V.

1978-01-01

Pressure tube 2566 was removed from N Reactor in July, 1977 to initiate the postirradiation examination program required by the Technical Specifications. Destructive examination of the pressure tube, after a maximum accumulated fluence of 4.6 x 10 21 n/cm 2 (E > 1 MeV), was conducted at the Hanford Engineering Development Laboratory to determine the effects of reactor service on the mechanical properties and hydrogen absorption and corrosion characteristics of the pressure tube. Tube 2566 is the sixth tube removed for destructive examination since the initial reactor startup. Evaluation of test results reveal that no significant detrimental changes have occurred in the parameters studied, since the last tube was removed in 1974

The effect of texture on delayed hydride cracking in Zr-2.5Nb alloy

Energy Technology Data Exchange (ETDEWEB)

Resta Levi, R.; Sagat, S

1999-09-01

Pressure tubes for CANDU reactors are made of Zr-2.5Nb alloy. They are produced by hot extrusion followed by cold work, which results in a material with a pronounced crystallographic texture with basal plane normals of its hexagonal structure around the circumferential direction. Under certain conditions, this material is susceptible to a cracking mechanism called delayed hydride cracking (DHC). Our work investigated the susceptibility of Zr-2.5Nb alloy pressure tube to DHC in this pressure tube material, in terms of crystallographic texture and grain shape. The results are presented in terms of crack velocity obtained on different planes and directions of the pressure tube. The results show that it is more difficult for a crack to propagate at right angles to crystallographic basal planes (which are close to the precipitation habit plane of hydrides) than for it to propagate parallel to the basal plane. However, if the cracking plane is oriented parallel to preexisting hydrides (hydrides formed as a result of the manufacturing process), the crack propagates along these hydrides easily, even if the hydride habit planes are not oriented favourably. (author)

Modelling flow and work hardening behaviour of cold worked Zr–2.5Nb pressure tube material in the temperature range of 30–600 oC

International Nuclear Information System (INIS)

Dureja, A.K.; Sinha, S.K.; Pawaskar, D.N.; Seshu, P.; Chakravartty, J.K.; Sinha, R.K.

2014-01-01

Under a postulated accident scenario of loss of cooling medium in an Indian Pressurised Heavy Water Reactor (IPHWR), temperature of the pressure tubes can rise and lead to large deformations. In order to investigate the modes of deformation of pressure tube – calandria tube assembly, material property data defining the flow behaviour over a temperature range from room temperature (RT) to 800 o C are needed. It is of practical importance to formulate mathematical equations to describe the stress–strain relationships of a material for a variety of reasons, such as the analysis of forming operations and the assessment of component's performance in service. A number of constitutive relations of empirical nature have been proposed and they have been found very suitable to describe the behaviour of a material. Although these relations are of empirical nature, various metallurgical factors appear to decide applicability of each of these relations. For example, grain size influences mainly the friction stress while the strain hardening is governed by dislocation density. In a recent work, tensile deformation behaviour of pressure tube material of IPHWR has been carried out over a range of temperature and strain rates (Dureja et al., 2011). It has been found that the strength parameters (yield and ultimate tensile strength) vary along the length of the tube with higher strength at the trailing end as compared to the leading end. This stems from cooling of the billet during the extrusion process which results in the variation of microstructure, texture and dislocation density from the leading to the trailing end. In addition, the variation in metallurgical parameters is also expected to influence the work hardening behaviour, which is known to control the plastic instability (related to uniform strain). In the present investigation, the tensile flow and work-hardening behaviour of a cold worked Zr–2.5Nb pressure tube material of IPHWRs has been studied over the

Modelling of pressure tube Quench using PDETWO

International Nuclear Information System (INIS)

Parlatan, Y.; Lei, Q.M.; Kwee, M.

2004-01-01

Transient two-dimensional heat conduction calculations have been carried out to determine the time-dependent temperature distribution in an overheated pressure tube during quenching with water. The purpose of the calculations is to provide input for evaluation of thermal (secondary) stresses in the pressure tube due to quench. The quench phenomenon in pressure tubes could occur in several hypothetical accident scenarios, including incidents involving intermittent buoyancy-induced flow during outages. In these scenarios, there will be two (radial and axial) or three dimensional temperature gradients, resulting in thermal stresses in the pressure tube, as the water front reaches and starts to cool down the hot pressure tube. The transient, two-dimensional heat conduction equation in the pressure tube during quench is solved using a FORTRAN package called PDETWO, available in the open literature for solving time-dependent coupled systems of non-linear partial differential equations over a two-dimensional rectangular region. This routine is based on finite difference solution of coupled, non-linear partial differential equations. Temperature gradient in the circumferential gradient is neglected for conservatism and convenience. The advancing water front is not modelled explicitly, and assumed to be at a uniform temperature and moving at a constant velocity inferred from experimental data. For outer surface and both ends of the pressure tube in the axial direction, a zero-heat flux boundary condition is assumed, while for the inner surface a moving water-quench front is assumed by appropriately varying the fluid temperature and the heat transfer coefficient. The pressure tube is assumed to be at a uniform temperature of 400 o C initially, to represent conditions expected during an intermittent buoyancy-influenced flow scenario. The results confirm the expectations that axial temperature gradients and associated heat fluxes are small in comparison with those in the

Pressure vessels and methods of sealing leaky tubes disposed in pressure vessels

International Nuclear Information System (INIS)

Larson, G.C.

1980-01-01

This invention relates to pressure vessels and to methods of sealing leaky tubes in them and is especially applicable to pressure vessels in the form of sheet-and-tube type heat exchangers constructed with a large number of relatively small diameter tubes grouped in a bundle. To seal off a leaky tube in such a heat exchanger an explosive activated plug in the form of a hollow metal body is used, inserted at each end of the tube to be sealed. Using the arrangement of pressure vessel and associated tube sheets and the explosive activated plug method of sealing a leaky tube as described in this invention it is claimed that distortion of the adjacent tubes and the tube sheets is reduced when the explosive activated plugs are detonated. (U.K.)

Creep deformation behaviour and microstructural changes in Zr-2.5% Nb alloy

International Nuclear Information System (INIS)

Chaudhuri, S.; Singh, R.; Ghosh, R.N.; Sinha, T.K.; Banerjee, S.

2002-01-01

Cold worked and stress relieved Zr-2.5% Nb alloy is a well-known material used as pressure tubes in Pressurised Heavy Water Reactors. The pressure tubes, made of a typical Zr-alloy, consisting of 2.54% Nb, 0.1175% oxygen and less than 100 ppm impurities, are expected to withstand 9.5 MPa to 12.5 MPa pressure at 250 degC to 310 degC under fast neutron fluxes of 3.5 x 10 17 nm -2 s -1 . These tubes are made by hot extrusion at 780 degC with an extrusion ratio 8.3:1 and 40% cold pilgering followed by annealing at 550 degC for 3 hours and subsequently by 20-30% cold pilgering and stress relieving at 400 degC for 24 hours. The microstructure of such cold worked and stress relieved alloy consists of Β-Zr precipitates in the matrix of elongated Α-Zr grains. Although various factors such as irradiation creep, thermal creep, irradiation growth etc are responsible for limiting the life of pressure tubes; the thermal creep contributes significantly in overall creep deformation. Keeping this in view as well as due to non-availability of adequate published information including creep database on this alloy, an extensive investigation on the thermal creep behaviour of indigenously produced Zr-2.5% Nb alloy was undertaken. The creep tests in air using Mayes' creep testing machines were carried out in the temperature range of 300 degC to 450 degC under stresses in the range of 50 to 550 MPa. Analysis of data revealed that the mechanism of creep deformation remains the same in this range

Investigation of Bragg reflections in α NbDsub(x) under hydrostatic pressure by γ ray diffraction

International Nuclear Information System (INIS)

Blaschko, O.; Klemencic, R.; Weinzierl, P.; Eder, O.J.

1978-01-01

Bragg reflections of NbDsub(x) single crystals in the α phase were studied under hydrostatic pressure using γ ray diffraction. The integrated intensity of the (211) reflection decreases by - 1.5 +- 0.2%/kbar and - 0.7 +- 0.1%/kbar in NbDsub(0.04) and NbDsub(0.02) respectively. The (200), (110) and (222) reflections show no change of intensity under hydrostatic pressure. In a pure Nb crystal no decrease of the intensity of the (211) reflection was found under hydrostatic pressure. (author)

Pressure tube rupture in a closed tank

International Nuclear Information System (INIS)

Khater, H.A.; Hadaller, G.I.; Stern, F.

1985-06-01

A study has been prepared on the feasibility of conducting pressure tube/calandria tube rupture tests in a closed tank, simulating a scaled-down calandria vessel. The study includes: i) a review of previous work, ii) an analytical investigation of the scaling problem of the calandria vessel and relevant in-core structures, iii) selection of a method for initiating pressure tube/calandria tube rupture, iv) a set of specifications for the test assembly, v) general arrangement drawings, vi) a proposal for a test matrix, vii) a survey and evaluation of existing facilities which could provide the required high pressure, temperature and fluid inventory, and viii) a cost estimate for the detailed design and construction, instrumentation, data acquisition and reduction, testing and reporting. The study concludes that it is both technically and practically feasible to conduct pressure tube rupture tests in a closed tank

21 CFR 868.5860 - Pressure tubing and accessories.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Pressure tubing and accessories. 868.5860 Section... (CONTINUED) MEDICAL DEVICES ANESTHESIOLOGY DEVICES Therapeutic Devices § 868.5860 Pressure tubing and accessories. (a) Identification. Pressure tubing and accessories are flexible or rigid devices intended to...

Resistivity of Weyl semimetals NbP and TaP under pressure

International Nuclear Information System (INIS)

Einaga, M.; Shimizu, K.; Hu, J.; Mao, Z.Q.; Politano, A.

2017-01-01

The resistivity of Weyl semimetals NbP and TaP has been investigated as a function of pressure and temperature. The behaviour of the resistivity as a function of pressure and temperature is closely correlated to the location of the Weyl points compared to the Fermi energy. The rapid increase of the resistivity in TaP and NbP under the application of 4.5 and 8.0 GPa is related with the shift of Weyl points, which affords a finite density of states near the Fermi energy. Specifically, we find that under pressure the Weyl points are situated above the Fermi energy. As regards the temperature behaviour, we detect a nonmonotonous behaviour of resistivity in TaP at 8.7 and 9.8 GPa as a function of temperature, whereas in the case of NbP the behaviour is more complicate. (copyright 2017 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Resistivity of Weyl semimetals NbP and TaP under pressure

Energy Technology Data Exchange (ETDEWEB)

Einaga, M.; Shimizu, K. [KYOKUGEN, Graduate School of Engineering Science, Osaka University, Toyonaka (Japan); Hu, J.; Mao, Z.Q. [Department of Physics and Engineering Physics, Tulane University, New Orleans, LA (United States); Politano, A. [Fondazione Istituto Italiano di Tecnologia, Graphene Labs, Genova (Italy)

2017-08-15

The resistivity of Weyl semimetals NbP and TaP has been investigated as a function of pressure and temperature. The behaviour of the resistivity as a function of pressure and temperature is closely correlated to the location of the Weyl points compared to the Fermi energy. The rapid increase of the resistivity in TaP and NbP under the application of 4.5 and 8.0 GPa is related with the shift of Weyl points, which affords a finite density of states near the Fermi energy. Specifically, we find that under pressure the Weyl points are situated above the Fermi energy. As regards the temperature behaviour, we detect a nonmonotonous behaviour of resistivity in TaP at 8.7 and 9.8 GPa as a function of temperature, whereas in the case of NbP the behaviour is more complicate. (copyright 2017 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

First-principles predictions of structural, mechanical and electronic properties of βTiNb under high pressure

Science.gov (United States)

Wang, Z. P.; Fang, Q. H.; Li, J.; Liu, B.

2018-04-01

Structural, mechanical and electronic properties of βTiNb alloy under high pressure have been investigated based on the density functional theory (DFT). The dependences of dimensionless volume ratio, elastic constants, bulk modulus, Young's modulus, shear modulus, ductile/brittle, anisotropy and Poisson's ratio on applied pressure are all calculated successfully. The results reveal that βTiNb alloy is mechanically stable under pressure below 23.45 GPa, and the pressure-induced phase transformation could occur beyond this critical value. Meanwhile, the applied pressure can effectively promote the mechanical properties of βTiNb alloy, including the resistances to volume change, elastic deformation and shear deformation, as well as the material ductility and metallicity. Furthermore, the calculated electronic structures testify that βTiNb alloy performs the metallicity and the higher pressure reduces the structural stability of unit cell.

The Study of Heat Treatment Effects on Chromium Carbide Precipitation of 35Cr-45Ni-Nb Alloy for Repairing Furnace Tubes

Directory of Open Access Journals (Sweden)

Nakarin Srisuwan

2016-01-01

Full Text Available This paper presents a specific kind of failure in ethylene pyrolysis furnace tubes. It considers the case in which the tubes made of 35Cr-45Ni-Nb high temperature alloy failed to carburization, causing creep damage. The investigation found that used tubes became difficult to weld repair due to internal carburized layers of the tube. The microstructure and geochemical component of crystallized carbide at grain boundary of tube specimens were characterized by X-ray diffractometer (XRD, scanning electron microscopy (SEM with back-scattered electrons mode (BSE, and energy dispersive X-ray spectroscopy (EDS. Micro-hardness tests was performed to determine the hardness of the matrix and the compounds of new and used tube material. The testing result indicated that used tubes exhibited a higher hardness and higher degree of carburization compared to those of new tubes. The microstructure of used tubes also revealed coarse chromium carbide precipitation and a continuous carbide lattice at austenite grain boundaries. However, thermal heat treatment applied for developing tube weld repair could result in dissolving or breaking up chromium carbide with a decrease in hardness value. This procedure is recommended to improve the weldability of the 35Cr-45Ni-Nb used tubes alloy.

Pressure tube reactors

International Nuclear Information System (INIS)

Natori, Hisahide.

1981-01-01

Purpose: To improve the electrical power generation efficiency in a pressure tube reactor in which coolants and moderators are separated by feedwater heating with heat generated in heavy water and by decreasing the amount of steams to be extracted from the turbine. Constitution: A heat exchanger and a heavy water cooler are additionally provided to a conventional pressure tube reactor. The heat exchanger is disposed at the pre-stage of a low pressure feedwater heater series. High temperature heavy water heated in the core is passed through the primary side of the exchanger, while feedwater is passed through the secondary side. The cooler is disposed on the downstream of the heat exchanger in the flowing direction of the heavy water, in which heavy water from the heat exchanger is passed through the primary side and the auxiliary equipment cooling water is sent to the secondary side thereof. Accordingly, since extraction of heating steams is no more necessary, the steam can be used for the rotation of the turbine, and the electrical power generation efficiency can be improved. (Seki, T.)

Powder-in-tube (PIT) Nb$_{3}$Sn conductors for high-field magnets

CERN Document Server

Lindenhovius, J L H; den Ouden, A; Wessel, W A J; ten Kate, H H J

2000-01-01

New Nb/sub 3/Sn conductors, based on the powder-in-tube (PIT) process, have been developed for application in accelerator magnets and high-field solenoids. For application in accelerator magnets, SMI has developed a binary 504 filament PIT conductor by optimizing the manufacturing process and adjustment of the conductor lay-out. It uniquely combines a non-copper current density of 2680 A/mm/sup 2/@10 T with an effective filament diameter of about 20 mu m. This binary conductor may be used in a 10 T, wide bore model separator dipole magnet for the LHC, which is being developed by a collaboration of the University of Twente and CERN. A ternary (Nb/7.5wt%Ta)/sub 3/Sn conductor containing 37 filaments is particularly suited for application in extremely high-field superconducting solenoids. This wire features a copper content of 43%, a non-copper current density of 217 A/mm/sup 2/@20 T and a B/sub c2/ of 25.6 T. The main issues and the experimental results of the development program of PIT Nb/sub 3/Sn conductors a...

Carbides precipitated from the melt in a Zr-2.5 Nb alloy

International Nuclear Information System (INIS)

Piotrkowski, R.; Garcia, E.A.; Vigna, G.L.; Bermudez, S.E.

1993-01-01

An experimental method is presented which leads to the formation of carbides similar in size (3 to 8 microns) and composition to those observed in some pressure tubes of CANDU type reactors. The method is based on melting the Zr-2.5 Nb alloy in a graphite crucible, where isothermal C diffusion in the Zr-Nb melt took place. It can be inferred that the carbides observed in pressure tubes could be originated in high temperature stages of the manufacture process. Otherwise, they could have been incorporated in the Zr sponge. As a result of the diffusion couple Liquid Zr-2.5 Nb/Solid Graphite, a carbide layer, up to 100μm thick, grew attached to the crucible wall, together with carbide particles whose size was in the some microns range. The smallest particles were arranged in rows determined by the prior β phase grains. The main carbide phase detected was the cubic MC 1-x ; the hexagonal M 2 C was also detected; M for metal. (Author)

Leak-before-break concept for evaluation of flows detected in pressure tubes in a Candu type reactor

International Nuclear Information System (INIS)

Crespi, J.C.

1992-01-01

This paper reviews the role of the Leak-Before-Break concept for evaluation of flaws detected in cold-worked Zr 2.5% Nb pressure tubes in a CANDU type reactors. The acceptance criteria are intended to prevent failure by brittle fracture, plastic collapse of the ligament and delayed hydride cracking. The methodology developed here was of great help in order to establish the operative conditions for fuel channel garter springs repositioning by means of the SLA Rette tool at Embalse Nuclear Generating Station, Cordoba, Argentina. (author)

Mechanism and deuterium pickup in Zr-2.5Nb alloy

International Nuclear Information System (INIS)

Ploc, R.A.

1999-12-01

There are approximately 400 Zr-2.5Nb pressure tubes in a CANDU reactor. During operation, the pressure tubes contain heavy water at about 300 deg C, 10.3 NPa with a room-temperature pD of 10.5. Operation of the pressure tube in the environment leads to oxide formation and absorption of deuterium. Excess deuterium absorption leads to precipitation of zirconium deuterides in the metal. A knowledge of how the deuterium passes through the oxide film to enter into the metal is an important step in gaining control over ingress rates. Fresnel fringe imaging of cross-sectioned oxides grown on pressure tubes, combined with tilting in the electron microscope, has revealed the three-dimensional nature of porosity in the oxide films. Two primary types exist, flake and ribbon. The main route for deuterium ingress is via ribbon porosity, as shown by electrochemical impedance spectroscopy. The location of the ribbon porosity is along the boundary between the oxidized α-Zr and β-Zr phases. Modifications to reduce ribbon porosity are possible and this, in turn, leads to significantly lower rates of deuterium absorption and extension of pressure-tube lifetime. (author)

Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

Energy Technology Data Exchange (ETDEWEB)

Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

2012-02-15

Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

Probabilistic fracture mechanics applied for DHC assessment in the cool-down transients for CANDU pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Radu, Vasile, E-mail: vasile.radu@nuclear.ro [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania); Roth, Maria [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania)

2012-12-15

For CANDU pressure tubes made from Zr-2.5%Nb alloy, the mechanism called delayed hydride cracking (DHC) is widely recognized as main mechanism responsible for crack initiation and propagation in the pipe wall. Generation of some blunt flaws at the inner pressure tube surface during refueling by fuel bundle bearing pad or by debris fretting, combined with hydrogen/deuterium up-take (20-40 ppm) from normal corrosion process with coolant, may lead to crack initiation and growth. The process is governed by hydrogen hysteresis of terminal solid solubility limits in Zirconium and the diffusion of hydrogen atoms in the stress gradient near to a stress spot (flaw). Creep and irradiation growth under normal operating conditions promote the specific mechanisms for Zirconium alloys, which result in circumferential expansion, accompanied by wall thinning and length increasing. These complicate damage mechanisms in the case of CANDU pressure tubes that are also are affected by irradiation environment in the reactor core. The structural integrity assessment of CANDU fuel channels is based on the technical requirements and methodology stated in the Canadian Standard N285.8. Usually it works with fracture mechanics principles in a deterministic manner. However, there are inherent uncertainties from the in-service inspection, which are associated with those from material properties determination; therefore a necessary conservatism in deterministic evaluation should be used. Probabilistic approach, based on fracture mechanics principle and appropriate limit state functions defined as fracture criteria, appears as a promising complementary way to evaluate structural integrity of CANDU pressure tubes. To perform this, one has to account for the uncertainties that are associated with the main parameters for pressure tube assessment, such as: flaws distribution and sizing, initial hydrogen concentration, fracture toughness, DHC rate and dimensional changes induced by long term

Zirconium hydrides and Fe redistribution in Zr-2.5%Nb alloy under ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Idrees, Y.; Yao, Z. [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, ON, Canada, K7L 3N6 (Canada); Cui, J.; Shek, G.K. [Kinetrics, Mississauga, ON (Canada); Daymond, M.R., E-mail: daymond@queensu.ca [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, ON, Canada, K7L 3N6 (Canada)

2016-11-15

Zr-2.5%Nb alloy is used to fabricate the pressure tubes of the CANDU reactor. The pressure tube is the primary pressure boundary for coolant in the CANDU design and is susceptible to delayed hydride cracking, reduction in fracture toughness upon hydride precipitation and potentially hydride blister formation. The morphology and nature of hydrides in Zr-2.5%Nb with 100 wppm hydrogen has been investigated using transmission electron microscopy. The effect of hydrides on heavy ion irradiation induced decomposition of the β phase has been reported. STEM-EDX mapping was employed to investigate the distribution of alloying elements. The results show that hydrides are present in the form of stacks of different sizes, with length scales from nano- to micro-meters. Heavy ion irradiation experiments at 250 °C on as-received and hydrided Zr-2.5%Nb alloy, show interesting effects of hydrogen on the irradiation induced redistribution of Fe. It was found that Fe is widely redistributed from the β phase into the α phase in the as-received material, however, the loss of Fe from the β phase and subsequent precipitation is retarded in the hydrided material. This preliminary work will further the current understanding of microstructural evolution of Zr based alloys in the presence of hydrogen. - Graphical abstract: STEM HAADF micrographs at low magnification showing the hydride structure in Zr-2.5Nb alloy.

Effect of Nb additions on the microstructure, thermal stability and mechanical behavior of high pressure Zr phases under ambient conditions

International Nuclear Information System (INIS)

Zhilyaev, A.P.; Sabirov, I.; Gonzalez-Doncel, G.; Molina-Aldareguia, J.; Srinivasarao, B.; Perez-Prado, M.T.

2011-01-01

Research highlights: → We analyze the influence of Nb additions on the shear-induced α → ω → β phase transformations in pure Zr by high pressure torsion (HPT). → Nb reduces the transition pressures and increases the transformation kinetics. → High pressure phases are retained under ambient conditions due to the presence of an internal stress. → Post-HPT annealing allows to fabricate bimodal/biphase nanostructures with enhanced mechanical behavior. - Abstract: This paper analyzes the influence of Nb on the shear-induced α → ω → β transformation taking place when processing Zr by high pressure torsion (HPT) under suitable conditions of pressure and shear. With that purpose, pure Zr and Zr-2.5%Nb were processed by HPT at room temperature and at pressures ranging from 0.25 to 6 GPa using 5 anvil turns. Nb causes a further reduction of the transition pressures, which are already lower when applying shear besides pressure. Thus, the transition pressure to the β phase is reduced at least 100 times in the Zr-Nb alloy. Alloying with Nb decreases the grain size of the transformed phases, significantly enhances their thermal stability and increases their UTS and elongation to failure. Selected post-HPT annealing treatments lead to the development of very tough, multiphase Zr and Zr-Nb with bimodal grain size distributions. The retention of the high pressure phases under ambient conditions is explained by the development of a high internal stress during processing. This stress is measured by synchrotron radiation diffraction at HZB-BESSY II. It is proposed that the presence of Nb reduces the internal stress level required for the retention of the high pressure phases.

In situ sampling for pressure tube deuterium concentration

International Nuclear Information System (INIS)

Harrington, A.J.; Kittmer, C.A.

1988-01-01

The present method of assessing the useful life of pressure tubes in CANDU (CANada Deuterium Uranium) reactors requires the periodic removal and examination of a tube. Special tooling was developed at Atomic Energy of Canada Limited (AECL) to obtain a sample of material from a pressure tube without removing the tube from the reactor. The sampling tool concept has been successfully used by Ontario Hydro during scheduled outages at the Pickering Nuclear Generating Station (PNGS). (author)

Corrosion of electron-irradiated Zr-2.5Nb and Zircaloy-2

International Nuclear Information System (INIS)

Woo, O.-T.; McDougall, G.M.; Hutcheon, R.M.; Urbanic, V.F.; Griffiths, M.; Coleman, C.E.

2000-01-01

We used 10-MeV electrons to rapidly produce radiation damage in zirconium alloys, investigated whether electrons produced the same microstructural changes as neutrons, then performed post-irradiation corrosion tests to determine whether electron-irradiated materials displayed similar corrosion behavior to neutron-irradiated materials. Two irradiations were completed using 10-MeV electrons with the beam normal to thin disks of material of 4 diameter slightly larger than the beam. The beam distribution. and disk cooling were designed to produce radial temperature and dose distributions having maxima at the disk center. A high-temperature irradiation was performed on annealed Zr-2.5Nb disks, achieving a central dose of 1.3 dpa and at a central temperature of ∼450 deg C. After irradiation, the samples contained needle-like β-Nb precipitates in the α-Zr matrix similar to those produced by neutrons. A low-temperature irradiation was performed on half-moon disks of Zr-2.5Nb and Zircaloy-2 pressure tube materials at 310 deg C central temperature and 1.3-dpa central dose. Dislocation loops were observed, again similar to those produced in neutron-irradiated materials. Some of the high-temperature electron-irradiated disks were exposed to 300 deg C moist air (saturated with D 2 O), and in separate tests, high- and low-temperature irradiated disks were corroded in 300 deg C D 2 0 (11.0 pD at room temperature) in an autoclave. Measurements of oxide thickness by Fourier Transform Infrared Reflectance (FTIR) spectroscopy showed that electron irradiation reduced the corrosion rate of Zr-2.5Nb compared with that of unirradiated material, as observed for neutron irradiation. For exposures to moist air and to D 2 O, the theoretical deuterium uptakes for the electron-irradiated materials were, respectively, about 4 times and 1.5 to 2 times those for the unirradiated materials. This is also in good agreement with results for neutron-irradiated pressure tube materials. Thus, 10-Me

Rejection index for pressure tubes

International Nuclear Information System (INIS)

Mitchell, A.B.; Meneley, D.

1989-10-01

The objective of the present study was to establish a set of criteria (or Rejection Index) which could be used to decide whether a zirconium-2 1/2 w/o niobium pressure tube in a CANDU reactor should be removed from service due to in-service degradation. A critique of key issues associated with establishing a realistic rejection index was prepared. Areas of uncertainty in available information were identified and recommendations for further analysis and laboratory testing made. A Rejection Index based on the following limits has been recommended: 1) Limits related to design intent and normal operation: any garter spring must remain within the tolerance band specified for its design location; the annulus gas system must normally be operated in a circulating mode with a procedure in place for purging to prevent accumulation of deuterium. It must remain sensitive to leaks into any part of the systems; and pressure tube dimensions and distortions must be limited to maintain the fuel channels within the original design intent; 2) Limits related to defect tolerance: adequate time margins between occurrence of a leaking crack and unstable failure must be demonstrated for all fuel channels; long lap-type flaws are unacceptable; crack-like defects of any size are unacceptable; and score marks, frat marks and other defects with contoured profiles must fall below certain depth, length and stress intensity limits; and 3) Limits related to property degradation: at operating temperature each pressure tube must be demonstrated to have a critical length in excess of a stipulated value; the maximum equivalent hydrogen level in any pressure tube should not exceed a limit which should be defined taking into account the known history of that tube; the maximum equivalent hydrogen level in any rolled joint should not exceed a limit which is presently recommended as 200 ppm equivalent hydrogen; and the maximum diametral creep strain should be limited to less than 5%

The role of strain localization in the fracture of irradiated pressure tube material

International Nuclear Information System (INIS)

Dutton, R.

1989-04-01

This report reviews those phenomena that lead to strain localization in zirconium alloys, with particular reference to the role played by the formation of shear bands in fracture processes. The important influence of plastic deformation, in general, on fracture mechanisms is emphasized. This is to be expected when elastic-plastic fracture mechanics is the chosen analytical technique. Intensely inhomogeneous characteristics of strain localization cause an abrupt bifurcation in the evolution of deformation strain and lead to plastic instability linked with intrinsic material behaviour (e.g., work softening) or of geometric origin (e.g., localized necking). Both of these effects are discussed in relation to measurable deformation parameters, such as the work hardening rate and strain rate sensitivity, which determine the degree of resistance to plastic instability. The modifying effect of irradiation on these quantities is given specific attention, the appropriate literature pertaining to Zircaloy and Zr-2.5% Nb being reviewed. Recommendations are made for a combined experimental and theoretical program to characterize strain localization and reduced ductility in irradiated cold-worked Zr-2.5% Nb pressure tube material. The relationship between the deformation properties and the fracture behaviour is discussed

Bulging of pressure tubes at hot spots under LOCA conditions

International Nuclear Information System (INIS)

Manu, C.; Shewfelt, R.S.W.; Wright, A.C.D.; Aboud, R.; Lau, J.H.K.; Sanderson, D.B.

1996-01-01

During certain postulated loss-of-coolant accidents (LOCA) in a CANDU reactor, some fuel channels can become highly voided within a very short time. Although the pressure tubes are heated mainly by convection and thermal radiation during the LOCA transient, additional heat flow occurs through the bearing pads that are in contact with the pressure tribe. This contact can lead to local hot spots and associated thermal stresses in the pressure tube wall. The two factors that affects the behavior of the pressure tubes during LOCA conditions are the internal pressure and the local heating. Although the effect of internal pressure and of axially uniform temperature has been studied elsewhere, the effect of the local heating on the pressure tube behavior has not been modelled before. This paper shows that the bulging of a pressure tube at a hot spot is the result of the thermal stresses that are developed in a pressure tube during a LOCA transient. To isolate the local heating effect from the internal pressure, a series of single-effect experiments was performed. In these experiments, sections of a CANDU pressure tube were subjected to local heating only. The thermal profile and the local deformation were measured function of time. To quantify the effect of the thermal stresses on the bulging of pressure tubes at hot spots and to develop numerical tools that can predict such bulging, finite element analyses were performed rising the ABAQUS finite element computer code. Use of the measured thermal profiles in the ABAQUS finite element analysis, resulted in very good agreement between the predicted and measured displacements. (author)

Endotracheal tube cuff pressure monitoring during neurosurgery - Manual vs. automatic method

Directory of Open Access Journals (Sweden)

Mukul Kumar Jain

2011-01-01

Full Text Available Background: Inflation and assessment of the endotracheal tube cuff pressure is often not appreciated as a critical aspect of endotracheal intubation. Appropriate endotracheal tube cuff pressure, endotracheal intubation seals the airway to prevent aspiration and provides for positive-pressure ventilation without air leak. Materials and Methods: Correlations between manual methods of assessing the pressure by an experienced anesthesiologists and assessment with maintenance of the pressure within the normal range by the automated pressure controller device were studied in 100 patients divided into two groups. In Group M, endotracheal tube cuff was inflated manually by a trained anesthesiologist and checked for its pressure hourly by cuff pressure monitor till the end of surgery. In Group C, endotracheal tube cuff was inflated by automated cuff pressure controller and pressure was maintained at 25-cm H 2 O throughout the surgeries. Repeated measure ANOVA was applied. Results: Repeated measure ANOVA results showed that average of endotracheal tube cuff pressure of 50 patients taken at seven different points is significantly different (F-value: 171.102, P-value: 0.000. Bonferroni correction test shows that average of endotracheal tube cuff pressure in all six groups are significantly different from constant group (P = 0.000. No case of laryngomalacia, tracheomalacia, tracheal stenosis, tracheoesophageal fistula or aspiration pneumonitis was observed. Conclusions: Endotracheal tube cuff pressure was significantly high when endotracheal tube cuff was inflated manually. The known complications of high endotracheal tube cuff pressure can be avoided if the cuff pressure controller device is used and manual methods cannot be relied upon for keeping the pressure within the recommended levels.

Frictional pressure drop of high pressure steam-water two-phase flow in internally helical ribbed tubes

International Nuclear Information System (INIS)

Tingkuan, C.; Xuanzheng, C.

1987-01-01

It is well known that the internally helical ribbed tubes are effective in suppressing the dry-out in boiling tubes at high pressures, so they are widely used as furnace water wall tubes in modern large steam power boilers. Design of the boilers requires the data on frictional pressure drop characteristics of the ribbed tubes, but they are not sufficient now. This paper describes the experimental results on the adiabatic frictional pressure drop in both horizontal ribbed tubes with measured mean inside diameter of 11.69 mm and 35.42 mm at high pressure from 10 to 21 MPa, mass flow rate from 350 to 3800 kg/m/sup 2/s and steam quality from 0 to 1 in our high pressure electrically heated water loop. Simultaneously, both smooth tubes under the same conditions for comparison. Based on the tests the correlation for determining the frictional pressure drop of internally ribbed tubes are proposed

Remotised sliver sample removal from irradiated pressure tube

Energy Technology Data Exchange (ETDEWEB)

Rupani, B B; Sharma, B S.V.G.; Shyam, T V; Sinha, R K [Bhabha Atomic Research Centre, Bombay (India). Reactor Engineering Div.

1994-12-31

The life of irradiated pressure tubes of Pressurised Heavy Water Reactor (PHWR) can be assessed by analysing hydrogen content in sliver samples of pressure tube material. The sample can be obtained in the form of small slivers by scraping at any specified locations within the bore of the pressure tube of operating reactor. A tool namely Hydride Scraping Tool (HST) has been developed to obtain very thin slivers of hydride bearing samples from irradiated pressure tube. In order to save man-rem consumption during scraping operation, a feeding mechanism has also been designed and developed for axial positioning of scraping tool in the channel remotely. This paper covers general details about constructional features of the scraping tool, feeding mechanism and their control system. It also highlights the operational salient features and capabilities of the system. Possible applications of the feeding mechanism in other fields are also indicated. (author). 4 figs.

Remotised sliver sample removal from irradiated pressure tube

International Nuclear Information System (INIS)

Rupani, B.B.; Sharma, B.S.V.G.; Shyam, T.V.; Sinha, R.K.

1994-01-01

The life of irradiated pressure tubes of Pressurised Heavy Water Reactor (PHWR) can be assessed by analysing hydrogen content in sliver samples of pressure tube material. The sample can be obtained in the form of small slivers by scraping at any specified locations within the bore of the pressure tube of operating reactor. A tool namely Hydride Scraping Tool (HST) has been developed to obtain very thin slivers of hydride bearing samples from irradiated pressure tube. In order to save man-rem consumption during scraping operation, a feeding mechanism has also been designed and developed for axial positioning of scraping tool in the channel remotely. This paper covers general details about constructional features of the scraping tool, feeding mechanism and their control system. It also highlights the operational salient features and capabilities of the system. Possible applications of the feeding mechanism in other fields are also indicated. (author). 4 figs

Pressure Tube and Pressure Vessel Reactors; certain comparisons

Energy Technology Data Exchange (ETDEWEB)

Margen, P H; Ahlstroem, P E; Pershagen, B

1961-04-15

In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D{sub 2}O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960.

Pressure Tube and Pressure Vessel Reactors; certain comparisons

International Nuclear Information System (INIS)

Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

1961-04-01

In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D 2 O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

Development of technology on the material surveillance of CANDU pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Noh, Kye Hoh; Han, Jung Hoh; Lee, Duk Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-05-01

Material degradation of pressure tubes, which are the most important components in CANDU fuel channel, can only be evaluated by removing and examining them(material surveillance). This study aimed at establishment of overall evaluation technology including the evaluation of the material degradation for the integrity of pressure tubes of Wolsung units. Material tests for pressure tubes were performed as follows; (1) Evaluation on life limiting factors of pressure tubes (2) Review on leak-before-break and integrity maintenance technology of pressure tubes (3) Survey on selection criteria for tubes to be inspected and on related regulations for material surveillance (4) Analysis of material surveillance test procedure (5) Basic examinations of Wolsung unit 1 pressure tube material(TEM, texture, chemical component etc) (6) Manufacture of test equipments and test (DHCV, hydriding, grip and tensile specimen etc). 23 figs, 6 tabs, 59 refs. (Author).

Development of technology on the material surveillance of CANDU pressure tubes

International Nuclear Information System (INIS)

Noh, Kye Hoh; Han, Jung Hoh; Lee, Duk Hyun

1995-05-01

Material degradation of pressure tubes, which are the most important components in CANDU fuel channel, can only be evaluated by removing and examining them(material surveillance). This study aimed at establishment of overall evaluation technology including the evaluation of the material degradation for the integrity of pressure tubes of Wolsung units. Material tests for pressure tubes were performed as follows; (1) Evaluation on life limiting factors of pressure tubes (2) Review on leak-before-break and integrity maintenance technology of pressure tubes (3) Survey on selection criteria for tubes to be inspected and on related regulations for material surveillance (4) Analysis of material surveillance test procedure (5) Basic examinations of Wolsung unit 1 pressure tube material(TEM, texture, chemical component etc) (6) Manufacture of test equipments and test (DHCV, hydriding, grip and tensile specimen etc). 23 figs, 6 tabs, 59 refs. (Author)

Tensile properties of Zr-2.5 Nb pressure tube alloy between 25 and 800 degC

International Nuclear Information System (INIS)

Singh, R.N.; Kishore, R.; Sinha, T.K.; Banerjee, S.

2000-10-01

Tensile properties of zirconium-2.5 wt. % niobium pressure tube material were evaluated by uniaxial tension tests at temperatures between 25 and 800 degC and under strain-rates varying from 3.3 x 10 -5 to 3.3 x 10 -3 /s. Tests were carried out on specimens fabricated from the sections of finished (autoclaved) tubes as well as on those machined from the sections of cold worked (2 nd pilgered) tubes. Moreover, specimens fabricated from finished tubes belonging to twenty different heats were tested at 300 degC to study the heat to heat variation in tensile properties of this alloy. In order to study the effect of the crystallographic texture on the tensile properties, specimens oriented in longitudinal as well as, in transverse directions of the tubes were also tested. Results showed that both yield and ultimate tensile strengths of this alloy decreased monotonically with increasing test temperatures, with a rapid fall in strengths above a temperature of 350 degC (623 K). The tensile ductility did not change appreciably up to 400 degC (673K) but increased rapidly above this temperature. The observed results on the temperature dependence of the strength and ductility indicated the possible occurrence of dynamic strain-ageing in this alloy in the temperature range of 200-300 degC (473 to 573 K). The transverse specimens showed higher strengths and lower ductility as compared to those of the longitudinal specimens up to a temperature of 350 degC (623 K). Above 350 degC, the difference in the strengths and the ductility of the two types of the specimens, became negligibly small indicating that the texture did not appreciably influence the tensile properties of this alloy at temperatures exceeding 350 degC. The alloy developed extensive superplasticity (ductility exceeding 100 %), when tested in the temperature range of 650-800 degC. Maximum ductility values of 650 % for longitudinal and 900 % for the transverse orientation with strain-rate sensitivity (m) exceeding 0

Advanced pressure tube sampling tools

International Nuclear Information System (INIS)

Wittich, K.C.; King, J.M.

2002-01-01

Deuterium concentration is an important parameter that must be assessed to evaluate the Fitness for service of CANDU pressure tubes. In-reactor pressure tube sampling allows accurate deuterium concentration assessment to be made without the expenses associated with fuel channel removal. This technology, which AECL has developed over the past fifteen years, has become the standard method for deuterium concentration assessment. AECL is developing a multi-head tool that would reduce in-reactor handling overhead by allowing one tool to sequentially sample at all four axial pressure tube locations before removal from the reactor. Four sets of independent cutting heads, like those on the existing sampling tools, facilitate this incorporating proven technology demonstrated in over 1400 in-reactor samples taken to date. The multi-head tool is delivered by AECL's Advanced Delivery Machine or other similar delivery machines. Further, AECL has developed an automated sample handling system that receives and processes the tool once out of the reactor. This system retrieves samples from the tool, dries, weighs and places them in labelled vials which are then directed into shielded shipping flasks. The multi-head wet sampling tool and the automated sample handling system are based on proven technology and offer continued savings and dose reduction to utilities in a competitive electricity market. (author)

Neutron radiography of irradiated zircaloy coupons of pressure tubes from PHWR`s

Energy Technology Data Exchange (ETDEWEB)

Gangotra, S; Ouseph, P M; Tamhane, A B; Singh, H N; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

1994-12-31

The Indian Pressurised Heavy Water Reactors (PHWR`s) are of CANDU type, consisting of 304 zircaloy-2 pressure tubes. These pressure tubes contain the fuel bundles, where the heat is generated and is removed by the heavy water flowing through these pressure tubes at high temperature and pressure. These pressure tubes are surrounded by the calandria tubes, and are separated from them by a pair of garter springs. Over a period of time, as a result of the irradiation creep and assisted by the displacement of the garter springs, the hot pressure tube may come in contact with the cold calandria tube. This would result in the hydrogen migrating to the cold contact location and formation of hydride blisters. These blisters could eventually rupture the pressure tube by the DHC (delayed hydrogen cracking) mechanism. 2 refs., 2 figs.

Pressure induced Amorphization of Ln1/3(Nb,Ta)O3

International Nuclear Information System (INIS)

Melchior, A.; Noked, O.; Sterer, E.; Shuker, R.

2014-01-01

The research focuses on the phenomenon of pressure induced amorphization (PIA) in Ln1/3MO3, Ln - La,Pr,Nd and M-Nb,Ta. In most pressure induced phase transitions the material changes from a crystalline phase to another crystalline phase. However, if this transition is kinetically hindered, the increased free energy due to the applied pressure will result in a structural collapse to an amorphous intermediate phase. This phenomenon is known as pressure induced amorphization

Innovation in pressure tube life assessment

International Nuclear Information System (INIS)

Guler, B.; Kalenchuk, D.; Celovsky, A.

2003-01-01

The hydrogen equivalent concentration and the rate of hydrogen ingress (in particular, deuterium) in pressure tubes are important parameters that must be assessed to determine the fitness-for-service of CANDU reactors. This paper presents the latest refinement in a process referred to as 'Pressure Tube Sampling', which is the only fully qualified and proven method that allows accurate determination of both the hydrogen equivalent concentration and the rate deuterium ingress without performing an expensive fuel channel removal. Pressure Tube Sampling has evolved over the past fifteen years during which over 2,300 samples have been obtained from CANDU reactors around the globe. In-reactor sampling is the standard method for determining the hydrogen equivalent concentrations and deuterium ingress rates in CANDU reactors. Over the past fifteen years, continual improvements in the Pressure Tube Sampling process have resulted in: the capability to obtain circumferential and axial samples, reduced 'on-face' time, reduced cost, reduced dose to workers, and improved analysis accuracy. Most recently, the new Multi-Head Sampling Tool (MHST) has been developed that continues this trend by using one tool to sample at all four axial pressure tube locations in a single visit to the fuel channel, thereby further improving efficiency. In 2001 October, the MHST was successfully deployed at Wolsong 1 by AECL for Korea Hydro and Nuclear Power. The tool was delivered using their Advanced Delivery Machine (ADM) and a total of sixteen samples were obtained from four channels. A significant saving in time was achieved with a rate of one channel (four samples) being sampled every 2 1/2 hours. For a typical 10-channel campaign, this could equate to a 2 to 3 days time/saving, which is significant in terms of outage schedule, cost, and worker dose. This paper provides a description of some of the latest innovations, with specific details on site application, performance, and end results

Simulation and analysis of the thermal and deformation behaviour of `as-received` and `hydrided` pressure tubes used in the circumferential temperature distribution experiments (end of life/pressure tube behaviour)

Energy Technology Data Exchange (ETDEWEB)

Muir, W C; Bayoumi, M H [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

It is postulated that in-reactor pressure tubes may be subjected to radiation damage and dissolved deuterium which could change the pressure tube characteristics and lead to different behaviour than that of as-received pressure tubes under large LOCA (loss of coolant) conditions. A hydrided pressure tube was used to study the effect of dissolved hydrogen on thermal-mechanical behaviour. In the experiment, simulating an in-reactor (hydrided) pressure tube with circumferential differential temperature under boil-off conditions, the pressure tube ballooned into contact with the calandria tube. The pressure tube used in this experiment was hydrided in a furnace to a nominal value of 200 {mu}g/g dissolved hydrogen. This test was a repeat of the first supplementary boil-off test (S-5-1) which used an as-received pressure tube. The objective of this paper is to analyze the results obtained from the simulation of this Boil-Off test using the SMARTT computer code and to examine the effect of hydriding on the thermal and ballooning behaviour of the pressure tube by comparison with the results obtained from test S-5-1. A discussion of the results obtained from this comparison is presented together with an analysis of their application to the analysis of pressure tube behaviour in CANDU reactors. (author). 13 refs., 1 tab., 16 figs.

Simulation and analysis of the thermal and deformation behaviour of 'as-received' and 'hydrided' pressure tubes used in the circumferential temperature distribution experiments (end of life/pressure tube behaviour)

International Nuclear Information System (INIS)

Muir, W.C.; Bayoumi, M.H.

1995-01-01

It is postulated that in-reactor pressure tubes may be subjected to radiation damage and dissolved deuterium which could change the pressure tube characteristics and lead to different behaviour than that of as-received pressure tubes under large LOCA (loss of coolant) conditions. A hydrided pressure tube was used to study the effect of dissolved hydrogen on thermal-mechanical behaviour. In the experiment, simulating an in-reactor (hydrided) pressure tube with circumferential differential temperature under boil-off conditions, the pressure tube ballooned into contact with the calandria tube. The pressure tube used in this experiment was hydrided in a furnace to a nominal value of 200 μg/g dissolved hydrogen. This test was a repeat of the first supplementary boil-off test (S-5-1) which used an as-received pressure tube. The objective of this paper is to analyze the results obtained from the simulation of this Boil-Off test using the SMARTT computer code and to examine the effect of hydriding on the thermal and ballooning behaviour of the pressure tube by comparison with the results obtained from test S-5-1. A discussion of the results obtained from this comparison is presented together with an analysis of their application to the analysis of pressure tube behaviour in CANDU reactors. (author). 13 refs., 1 tab., 16 figs

Delayed hydrogen cracking of zirconium alloy pressure tubes

International Nuclear Information System (INIS)

Jackman, A.H.; Dunn, J.T.

1976-10-01

After several years of almost continuous service, Pickering Units 3 and 4 have both experienced long outages to replace cracked pressure tubes. This report summarizes the status of the investigation into the cause of the cracks as of May 1976. The basic cause of the cracking was the presence of very high residual tensile stresses in the pressure tubes due to improper rolling procedures. These residual stresses are being reduced to acceptable levels by local stress relieving techniques at Bruce G.S. and in future reactors improvements in rolling procedures and changes in pressure tube specifications will prevent a recurrence of this problem. (author)

Tube micro-fouling, boiling and steam pressure after chemical cleaning

International Nuclear Information System (INIS)

Hu, M.H.

1998-01-01

This paper presents steam pressure trends after chemical cleaning of steam generator tubes at four plants. The paper also presents tube fouling factor that serves as an objective parameter to assess tubing boiling conditions for understanding the steam pressure trend. Available water chemistry data helps substantiate the concept of tube micro-fouling, its effect on tubing boiling, and its impact on steam pressure. All four plants experienced a first mode of decreasing steam pressure in the post-cleaning operation. After 3 to 4 months of operation, the decreasing trend stopped for three plants and then restored to a pre-cleaning value or better. The fourth plant is soil in decreasing trend after 12 months of operation. Dissolved chemicals, such as silica, titanium can precipitate on tube surface. The precipitate micro-fouling can deactivate or eliminate boiling nucleation sites. Therefore, the first phase of the post-cleaning operation suffered a decrease in steam pressure or an increase in fouling factor. It appears that micro fouling by magnetite deposit can activate or create more bubble nucleation sites. Therefore, the magnetite deposit micro-fouling results in a decrease in fouling factor, and a recovery in steam pressure. Fully understanding the boiling characteristics of the tubing at brand new, fouled and cleaned conditions requires further study of tubing surface conditions. Such study should include boiling heat transfer tests and scanning electronic microscope examination. (author)

Hydrogen embrittlement of Zr-2.5Nb PT with temperature

International Nuclear Information System (INIS)

Oh, Dong Joon; Ahn, Sang Bok; Kim, Young Suk

2003-01-01

The aim of this study is to investigate the effect of hydrogen embrittlement of Zr-2.5Nb CANDU pressure tube. The tests were performed at three hydrogen contents for transverse tensile and CCT specimens while the test temperatures were changed (RT to 300 .deg. C). The specimens were directly machined from the tube retaining original curvature using electric discharge machine. Both the transverse tensile and the fracture toughness tests showed the hydrogen embrittlement clearly at RT but this phenomenon was disappeared while the test temperature arrived over 250 .deg. C

Database for Pressure Tube Diameter and Operation Data of Wolsong NPP

International Nuclear Information System (INIS)

Jung, Jong Yeob; Kim, W. Y.; Bae, J. H.; Park, J. H.

2010-12-01

Pressure tube of CANDU reactor which is a long cylindrical shape of its diameter about 10 cm and length of about 6m, can be expanded toward both radial and axial directions due to irradiation under the high pressure and temperature condition. As the irradiation period increases, the radial expansion due to creep of the pressure tube increases. The radial expansion of the pressure tube comes out the reduction of the coolability and it results in the power deration. The objectives of the current work is to establish the database for the measured diametral data of pressure tube and operational data from Wolsong NPP as a preliminary work of developing the prediction model for pressure tube diameter. In order to develop the database, measured data for total 86 channels were collected from Wolsong NPP 1, 2, 3 and 4 and analyzed. Based on the provided data, the operational conditions such as an axial temperature and a pressure of the channel and neutron fluxes were derived. All data were analysed to derive the correlation between the pressure tube diameter and the other operational parameters

The effect of minor alloying elements on oxidation and hydrogen pickup in Zr-2.5Nb

International Nuclear Information System (INIS)

Ploc, R.A.

2002-01-01

In CANDU reactors, fuel and coolant are contained in horizontal pressure tubes made of Zr-2.5 wt% Nb alloy. In the past decade, the effect of more than 20 impurity elements, in various concentrations, on oxidation and deuterium pickup (at 300 o C, pD = 10.5, Li 2 O) have been investigated in over 70 Zr-2.5Nb alloys. The studies were performed using non-consumable arc-melted alloy logs that were rolled and made into corrosion coupons and corroded in autoclaves. This study represents one of the largest collections of previously unpublished data on the effect of impurity elements on oxide film growth and deuterium pickup in a zirconium alloy. Elements such as Al, Ti, Mn, and Pt, to name but a few, were found to significantly accelerate the corrosion process. Some elements, such as tin, had a positive effect on oxidation (lowers the rate of oxide film development) and a negative effect on hydrogen pickup (increases pickup). Three parameters were important to the corrosion process, namely, microstructure, surface finish, and synergistic interactions between the impurity elements. The above studies culminated in two response surface analyses (RSA). The first was conducted on the effect of C and Fe on oxygen and deuterium pickup in Zr-2.5Nb drop castings corroded at 325 o C in CANDU conditions. The second study was performed in autoclaves at 300 o C on the affect of four impurity elements, C, Fe, Cr, and Si, in Zr-2.5Nb micro-tubes, which possess the same microstructure as full-size pressure tubes. The first RSA revealed a quadratic dependence of corrosion on C and Fe concentrations, with an optimum resistance at about 30 ppm (wt) C and 1100 ppm (wt) Fe. This has been partially confirmed by out-reactor corrosion of Zr-2.5Nb-Fe micro-pressure tubes. Trends in- and out-reactor were similar for oxidation but different in magnitude for deuterium pickup. There is no linear dependence on the Fe concentration in-reactor, implying that Fe and C form a complex. The second RSA

The effect of minor alloying elements on oxidation and hydrogen pickup in Zr-2.5Nb

International Nuclear Information System (INIS)

Ploc, R.A.

2002-01-01

In CANDU reactors, fuel and coolant are contained in horizontal pressure tubes made of Zr-2.5 wt% Nb alloy. In the past decade, the effect of more than 20 impurity elements, in various concentrations, on oxidation and deuterium pickup (at 300 deg C, pD = 10.5, Li 2 O) have been investigated in over 70 Zr-2.5Nb alloys. The studies were performed using non-consumable arc-melted alloy logs that were rolled and made into corrosion coupons and corroded in autoclaves. This study represents one of the largest collections of previously unpublished data on the effect of impurity elements on oxide film growth and deuterium pickup in a zirconium alloy. Elements such as Al, Ti, Mn, and Pt, to name but a few, were found to significantly accelerate the corrosion process. Some elements, such as tin, had a positive effect on oxidation (lowers the rate of oxide film development) and a negative effect on hydrogen pickup (increases pickup). Three parameters were important to the corrosion process, namely, microstructure, surface finish, and synergistic interactions between the impurity elements. The above studies culminated in two response surface analyses (RSA). The first was conducted on the effect of C and Fe on oxygen and deuterium pickup in Zr-2.5Nb drop castings corroded at 325 deg C in CANDU conditions. The second study was performed in autoclaves at 300 deg C on the affect of four impurity elements, C, Fe, Cr, and Si, in Zr-2.5Nb micro-tubes, which possess the same microstructure as full-size pressure tubes. The first RSA revealed a quadratic dependence of corrosion on C and Fe concentrations, with an optimum resistance at about 30 ppm (wt) C and 1100 ppm (wt) Fe. This has been partially confirmed by out-reactor corrosion of Zr-2.5Nb-Fe micro-pressure tubes. Trends in- and out-reactor were similar for oxidation but different in magnitude for deuterium pickup. There is no linear dependence on the Fe concentration in-reactor, implying that Fe and C form a complex. The second

Stabilization of a Nb3Sn persistent current switch

International Nuclear Information System (INIS)

Urata, M.; Maeda, H.; Nakayama, S.; Yoneda, E.; Oda, Y.; Kumano, T.; Aoki, N.; Tomisaki, T.; Kabashima, S.

1993-01-01

A 2000 A class Nb 3 Sn persistent current switch has been successfully fabricated in the Toshiba R and D Center. The Nb tube processed conductor with Cu-10 wt.% Ni matrix has been developed for the switch in the Showa Electric Wire and Cable Co. Ltd. The magnetic instability which was observed in the previous 35 Ω Nb 3 Sn persistent current switch was improved in the present switch. The problem of quench current degradation and flux jump on magnetization, emerged in the previous switch, were confirmed to be solved. In the fast ramp, however, the switch degrades from the calculated results assuming the self field ac loss. In the Nb 3 Sn reaction process, Sn in the bronze diffuses into the Nb tube, which decreases the switch resistance. It was observed by a computer aided micro analysis (CMA) that Ni in the CuNi matrix precipitated on the Nb tube, which slightly reduced the switch resistance. (orig.)

Age hardening of cold-worked Zr-2.5 wt% Nb pressure tube alloy

International Nuclear Information System (INIS)

Kishore, R.; Singh, R.N.; Dey, G.K.; Sinha, T.K.

1992-01-01

Specimens for hardness and tensile tests, machined from a cold-worked zirconium-2.5% niobium pressure tube, with their axes parallel to longitudinal and transverse directions, were aged for 1 hr. at 300-500 C. The age hardening behaviour was monitored by mechanical tests, electron-microscopy and x-ray diffraction. In addition a few studies were carried on longitudinal tension specimens subjected to prolonged ageing (100-1000 hrs) at 300 C. It was observed that the short-term (1 hour) thermal ageing of this material at 300-400 C caused an increase in both strength and hardness without affecting ductility. It appears that the observed age-hardening is due to precipitation hardening by a niobium-rich phase and softening by recovery of cold-work and that the phenomenon is influenced by crystallographic texture. Further it was noted that a prolonged ageing at 300 C upto 1000 hrs, did not cause any appreciable changes in strength and ductility of the material compared to those obtained by 1 hour ageing at the same temperature. (author). 11 refs., 3 figs., 2 tabs

Corrosion of electron-irradiated Zr-2.5Nb and Zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Woo, O.-T.; McDougall, G.M.; Hutcheon, R.M.; Urbanic, V.F.; Griffiths, M.; Coleman, C.E

2000-07-01

We used 10-MeV electrons to rapidly produce radiation damage in zirconium alloys, investigated whether electrons produced the same microstructural changes as neutrons, then performed post-irradiation corrosion tests to determine whether electron-irradiated materials displayed similar corrosion behavior to neutron-irradiated materials. Two irradiations were completed using 10-MeV electrons with the beam normal to thin disks of material of 4 diameter slightly larger than the beam. The beam distribution. and disk cooling were designed to produce radial temperature and dose distributions having maxima at the disk center. A high-temperature irradiation was performed on annealed Zr-2.5Nb disks, achieving a central dose of 1.3 dpa and at a central temperature of {approx}450 deg C. After irradiation, the samples contained needle-like {beta}-Nb precipitates in the {alpha}-Zr matrix similar to those produced by neutrons. A low-temperature irradiation was performed on half-moon disks of Zr-2.5Nb and Zircaloy-2 pressure tube materials at 310 deg C central temperature and 1.3-dpa central dose. Dislocation loops were observed, again similar to those produced in neutron-irradiated materials. Some of the high-temperature electron-irradiated disks were exposed to 300 deg C moist air (saturated with D{sub 2}O), and in separate tests, high- and low-temperature irradiated disks were corroded in 300 deg C D{sub 2}0 (11.0 pD at room temperature) in an autoclave. Measurements of oxide thickness by Fourier Transform Infrared Reflectance (FTIR) spectroscopy showed that electron irradiation reduced the corrosion rate of Zr-2.5Nb compared with that of unirradiated material, as observed for neutron irradiation. For exposures to moist air and to D{sub 2}O, the theoretical deuterium uptakes for the electron-irradiated materials were, respectively, about 4 times and 1.5 to 2 times those for the unirradiated materials. This is also in good agreement with results for neutron-irradiated pressure

Specific aspects in the manufacturing and operating of CANDU reactor pressure tubes (P/T)

International Nuclear Information System (INIS)

Muscaloiu, C.

1997-01-01

The CANDU reactor design is based on a number of individual P/T in which nuclear fuel bundles are located. P/T are required to be operated in an environment of elevated temperature (300 o C), internal pressure (10 Mpa), fast neutron flux (E>1 MeV) and heavy water. The most suitable material which can provide the desired neutron economy and still maintain its mechanical properties along with corrosion resistance is zirconium alloys Zr+ 2.5 % Nb with the following composition: niobium, 2.5 to 2.8 weight percent; oxygen, 1,000 to 1,300 ppm; zirconium + allowed impurities - balance. A total of 380 pressure tubes are installed into reactor. Each pressure tube is attached at each end to a stainless steel end fitting by means of a grooved, expanded joint. The installation works were performed by ANM Bucuresti, under the technical support of General Electric Canada. The integrity of P/T after installation was examined as follows: - the surface of the rolled area on unrolled internal surface extending 25 mm beyond rolled area was inspected for irregularities by means of a boroscope; - all pressure tubes were subjected to the helium leak test after F/C installation. During P/T operating life periodical inspections according to Canadian Standard CSA N285.4 are performed. The selection of the P/T for inspection is based either on particular properties or on the operating conditions of the fuel channel. The inspection consists in: a) Base Line Inspection within 2 years period commencing after 7,000 EFPH of operation which will include a volumetric inspection over P/T full length and measurements of P/T sag, ID, wall thickness and F/C bearing positions; b) Periodic Inspection in the same conditions plus material surveillance (on the four most significant indication P/T detected during the Base Line Inspection). The inspection will be performed on 14 selected P/T. (author)

TEM examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

Srivastava, D.; Tewari, R.; Dey, G.K.; Sharma, B.P.; Sah, D.N.; Banerjee, Suparna; Sahoo, K.C.

2005-09-01

In the present work, microstructure of the zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water RAPP-1. Reactor (PHWR) has been examined for the first time using transmission electron microscope (TEM). The samples were obtained from a zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit -1, for a period for 6.77 effective full power years (EFPYs) and expected to have a cumulative radiation damage of about 3 dpa. In this study irradiated microstructure has been characterized and compared it with the microstructure of the unirradiated pressure tube samples. The effect of irradiation on the hydriding behaviour is also studied. (author)

Comparison of evaluation method for planar flaw in pressure tube

International Nuclear Information System (INIS)

Choi, Sung Nam; Kim, Hyung Nam; Yoo, Hyun Joo; Hwang, Won Gul

2009-01-01

CSA N285.4-94 requires the periodic inservice inspection and surveillance of pressure tubes in operating CANDU nuclear power reactors. If the inspection results reveal a flaw exceeding the acceptance criteria of the Code, the flaw must be evaluated to determine if the pressure is acceptable for continued service. Currently, the flaw evaluation methodology and acceptance criteria specified in CSA N285.8-05, 'Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors'. The Code is applicable to zirconium alloy pressure tubes. The evaluation methodology for a crack-like flaw is similar to that of FFSG(Fitness For Service Guideline for Zirconium alloy pressure in operation CANDU) used now. The object of this paper is to address the fracture initiation and plastic collapse evaluation for the planar flaw as it applies to the pressure tube on Wolsong NPP.

Remote ultrasonic characterisation of an irradiated pressure tube from RAPS-II

Energy Technology Data Exchange (ETDEWEB)

Gangotra, S; Muralidhar, S; Raut, S D; Ouseph, P M; Ghosh, J K; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

1994-12-31

The Rajasthan Atomic Power Station Unit-2 (RAPS-2) has reached a stage of operation where the contacting pressure tubes are suspect to failure as a result of irradiation creep and displacement of the garter springs, the hot pressure tube coming in contact with the cold calandria tube. To study and assess the safety of these pressure tubes, two channels believed to be in contact with the calandria tubes, have been removed from the reactor for detailed full length post irradiation examination. Some of the test results are presented. 2 refs., 3 figs., 1 tab.

Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

International Nuclear Information System (INIS)

Cheong, Y. M.; Kim, Y. S.; Gong, U. S.; Kwon, S. C.; Kim, S. S.; Choo, K.N.

2000-09-01

When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described

Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Cheong, Y M; Kim, Y S; Gong, U S; Kwon, S C; Kim, S S; Choo, K N

2000-09-01

When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described.

Full length channel Pressure Tube sagging under completely voided full length pressure tube of an Indian PHWR

Energy Technology Data Exchange (ETDEWEB)

Negi, Sujay, E-mail: negi.sujay@gmail.com [Indian Institute of Technology, Roorkee 247667 (India); Kumar, Ravi, E-mail: ravikfme@gmail.com [Indian Institute of Technology, Roorkee 247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Bhabha Atomic Research Centre, Mumbai 400085 (India); Mukopadhyay, D., E-mail: dmukho@barc.gov.in [Bhabha Atomic Research Centre, Mumbai 400085 (India)

2017-03-15

Highlights: • At 16 kW/m input, thermal stability was attained at 595 °C, without PT-CT contact. • At 20 kW/m step input, PT-CT contact occurred at 637 °C near bottom-center of the tube. • PT integrity was maintained throughout the experiment. - Abstract: An experimental investigation was conducted to simulate the sagging behavior of a full length Pressure Tube of a channel of 220 MWe Indian PHWR. The investigation aimed to recreate a condition resembling Loss of Coolant Accident (LOCA) with Emergency Core Cooling System (ECCS) failure in a nuclear power plant. A full length channel assembly immersed in moderator was subjected to electrical resistance heating of Pressure Tube (PT) to simulate the residual heat after shutting down of reactor. The temperature of PT started rising and the contact between PT and CT was established at the center of the tube where average bottom temperature was 637 °C. The integrity of PT was maintained throughout the experiment and the PT heat up was arrested on contact with the CT due to transfer of heat to the moderator.

The cracking of pressure tubes in the Pickering reactor

International Nuclear Information System (INIS)

Ross-Ross, P.A.

1978-01-01

Small cracks in 17 of the 390 pressure tubes in Unit 3 of the 2056 MW (electrical) Pickering Generating Station and of 52 tubes in Unit 4, resulted in each of these units being out of service for many months. The cracks originated at areas of extremely high residual tensile stress produced by improper positioning of the rolling tool used during construction to join the pressure tube to its end-fitting. The mechanism of failure was delayed hydrogen cracking. (author)

Influence of microstructure on the accelerated corrosion in Zr-Nb alloys

International Nuclear Information System (INIS)

Muller, S; Lanzani, L

2012-01-01

The influence of microstructure on the accelerated corrosion of Zr-1%Nb and Zr-2.5%Nb (CANDU's pressure tube material) has been studied. The behavior of Zircaloy-4 was also studied in order to compare the Zr-Nb alloys with an alloy that does not have niobium as an alloying element. The corrosion tests were carried out in LiOH 0.1M at 340 o C, in LiOH 1M at the same temperature and in steam at 400 o C. The results showed that the behavior of Zr-Nb alloys in steam at 400 o C is similar to that of Zircaloy-4 in this medium. However, Zr-Nb alloys are more sensitive than Zircaloy-4 to the presence of LiOH. The results suggest that the niobium concentration in the matrix is the parameter that defines the oxidation rate in Zr-Nb alloys, while the presence of second phases in these alloys (β--Zr/β-Nb/Zr-Nb-Fe) could be related with the growth of non-protective oxides in LiOH solutions. In LiOH 1M, the corrosion resistance of Zr-Nb alloys is similar to that of Zircaloy-4, except for the Zr-1Nb martensitic material which showed a sharp increase in the oxidation rate in this medium (author)

Pressure distribution over tube surfaces of tube bundle subjected to two phase cross flow

International Nuclear Information System (INIS)

Sim, Woo Gun

2013-01-01

Two phase vapor liquid flows exist in many shell and tube heat exchangers such as condensers, evaporators and nuclear steam generators. To understand the fluid dynamic forces acting on a structure subjected to a two phase flow, it is essential to obtain detailed information about the characteristics of a two phase flow. The characteristics of a two phase flow and the flow parameters were introduced, and then, an experiment was performed to evaluate the pressure loss in the tube bundles and the fluid dynamic force acting on the cylinder owing to the pressure distribution. A two phase flow was pre mixed at the entrance of the test section, and the experiments were undertaken using a normal triangular array of cylinders subjected to a two phase cross flow. The pressure loss along the flow direction in the tube bundles was measured to calculate the two phase friction multiplier, and the multiplier was compared with the analytical value. Furthermore, the circular distributions of the pressure on the cylinders were measured. Based on the distribution and the fundamental theory of two phase flow, the effects of the void fraction and mass flux per unit area on the pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure on the tube by a numerical method. It was found that for low mass fluxes, the measured two phase friction multipliers agree well with the analytical results, and good agreement for the effect of the void fraction on the drag coefficients, as calculated by the measured pressure distributions, is shown qualitatively, as compared to the existing experimental results

Pressure-induced positive electrical resistivity coefficient in Ni-Nb-Zr-H glassy alloy

Science.gov (United States)

Fukuhara, M.; Gangli, C.; Matsubayashi, K.; Uwatoko, Y.

2012-06-01

Measurements under hydrostatic pressure of the electrical resistivity of (Ni0.36Nb0.24Zr0.40)100-xHx (x = 9.8, 11.5, and 14) glassy alloys have been made in the range of 0-8 GPa and 0.5-300 K. The resistivity of the (Ni0.36Nb0.24Zr0.40)86H14 alloy changed its sign from negative to positive under application of 2-8 GPa in the temperature range of 300-22 K, coming from electron-phonon interaction in the cluster structure under pressure, accompanied by deformation of the clusters. In temperature region below 22 K, the resistivity showed negative thermal coefficient resistance by Debye-Waller factor contribution, and superconductivity was observed at 1.5 K.

Use of CATHENA to model calandria-tube/moderator heat transfer after pressure-tube/calandria-tube ballooning contact

International Nuclear Information System (INIS)

Fan, H.Z.; Bilanovic, Z.; Nitheanandan, T.

2004-01-01

A study was performed to assess the effect of the calandria-tube/moderator heat transfer after pressure-tube/calandria tube ballooning contact using CATHENA. Results of this study indicated that the analytical tool, CATHENA, can be applied for pool boiling heat transfer on the external surface of a large diameter tube, such as the calandria tube used in CANDU reactors. The methodology in such CANDU-generic study can be used to simulate the tube surface with multiple boiling regimes and to assess the benefits of closely coupling thermalhydraulics modelling and fuel/fuel channel behaviour modelling. CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a one-dimensional, two-fluid thermalhydraulic simulation code designed by AECL to analyse two-phase flow and heat transfer in piping networks. The detailed heat transfer package in CATHENA allows a connection to be established from the multiple solid surfaces of tubes to the surrounding large amount of moderator water, which acts as a heat sink during a postulated loss of coolant event. The generalized heat transfer package within CATHENA allows the tube walls to be divided into several layers in the radial direction and several sectors in the circumferential direction, to account for heat transfer conditions in these two directions. The CATHENA code with the generalized heat transfer package is capable of capturing key pool-boiling phenomena such as nucleate, transition and film boiling heat transfer as well as an ability to model the rewet phenomenon to some extent. A CATHENA input model was generated and used in simulations of selected contact boiling experiment test cases. The transient wall temperatures have been calculated in different portions of the calandria tube. By using this model an adequate agreement was achieved between CATHENA calculation and experimental measurement The CATHENA code enables one to investigate the transient and local thermal-mechanical behaviour of the calandria tube

Consequences of pressure tube rupture on in-core components

International Nuclear Information System (INIS)

Hill, P.G.; Hauptmann, E.G.; Lee, V.

1982-12-01

An investigation has been made of the consequences of pressure tube rupture in calandria vessels of heavy water cooled and moderated reactors. The study included a review of previous experimental and analytical work, as well as supplementary investigations carried out to examine the validity of previous assumptions and findings. The central questions considered were: the possibility of a propagating pressure tube failure; damage to the calandria vessel; and damage to the shut-off-rod guide tubes of the reactor shut-down system. The results of the investigation do not indicate mechanisms of sufficient strength to cause propagating failure in a well-designed, well-operated reactor following a tube burst under normal operating conditions. However, not all the details of the physical processes involved in a tube burst have been revealed by existing experimental and analytical work

Experiment on the effects of contact between the pressure tube and the fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Y; Fujii, Y [Electric Power Development Co. Ltd., Tokyo (Japan); Kato, K [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

1996-12-31

The Advanced Thermal Reactor (ATR) is a pressure tube type reactor in which the fuel assembly is located close to the pressure tube. The ATR has a structure which is such that the thermal stretch of the fuel pin is not limited by the spacer if the fuel pin dries out. Accordingly. it is not thought that the fuel pin contacts the pressure tube due to large transformations around the Design Based Event (DBE). Nevertheless, the safety margin must be kept in case the over-DBE. We have confirmed in this experiment that the temperature of the pressure tube does not increase to the critical level when the fuel pin contacts the pressure tube and the functions of the pressure tube are maintained as a pressure boundary. Further, we analyzed the safety margin of the pressure tube using the data from this experiment and from code analysis. (author). 10 tabs., 32 figs.

Delayed hydride cracking velocity and crack growth measurement using DCPD technique in Zr-2.5Nb pressure tube material

International Nuclear Information System (INIS)

Singh, R.N.; Kishore, R.; Roychaudhury, S.; Unnikrishnan, M.; Sinha, T.K.; De, P.K.; Banerjee, S.; Kumar, Santosh

2000-12-01

Nuclear structural materials have to perform under most demanding and exotic environmental conditions. Due to its unique properties dilute zirconium alloys are the only choice for in-core structural materials in water cooled nuclear reactors. Hydrogen related problems have been recognized as the life-limiting factor for the core components of Pressurized Heavy Water Reactors (PHWR). Delayed Hydride Cracking (Dhc) is one of them. In this study, Dhc crack growth has been monitored using Direct Current Potential Drop (Dcp) technique. Calibration curve between normalized Dcp output and normalized crack length was established at different test temperatures. Dhc velocity was measured along the axial direction of the Zirconium-2.5Niobium pressure tube material at 203 and 250 degree C. (author)

X-ray diffraction line profile analysis for defect study in Zr–2⋅5% Nb ...

Indian Academy of Sciences (India)

Unknown

The microstructure characterization by X-ray line profile analysis is possible for determination of ... breadth method also gives the coherent domain size, dislocation density and micro-strain present in .... currently used for pressure tubes, has replaced Zircaloy-2 ..... the data on σ0, σNb, m, α, G, and b available in the litera-.

Assessment and management of ageing of major nuclear power plant components important to safety: CANDU pressure tubes

International Nuclear Information System (INIS)

1998-08-01

The report documents the current practices for assessment and management of the ageing of the pressure tubes in CANDU reactors and Indian PHWTRs. Chapter headings are: fuel channel and pressure tube description, design basis for the fuel channel and pressure tube, degradation mechanisms and ageing concerns for pressure tubes, inspection and monitoring methods for pressure tubes,assessment methods and fitness-for-service guidelines for pressure tubes, mitigation methods for pressure tubes, and pressure tube ageing management programme

Delayed Hydride Cracking in Zr-2.5Nb Tubes with the Direction of An Approach to Temperature

International Nuclear Information System (INIS)

Kim, Young Suk; Im, Kyung Soo; Kim, Kang Soo; Ahn, Sang Bok; Cheong, Yong Moo

2006-01-01

One of the unique features of delayed hydride cracking (DHC) of zirconium alloys is that the DHC velocity (DHCV) of zirconium alloys strongly depends on the path to the test temperature. Ambler reported that the DHCV of Zr-2.5Nb tubes at temperatures above 180 .deg. C depended upon the direction of an approach to the test temperatures, and reported on a presence of the DHC arrest temperature or TDAT above which the DHCV decreased upon an approach to the test temperature by a heating. Ambler proposed a hydrogen transfer from the bulk to the crack tip assuming that the hydrides formed at the crack tip and in the bulk region are fully constrained and partially constrained at the crack tip, respectively. In other words, the terminal solid solubility (TSS) of hydrogen would be governed by elastic strain energy induced by the precipitating hydrides, leading to a higher TSS in the bulk region than that at the crack tip. In a sense, his assumption that the hydrogen concentration is higher in the bulk region than that at the crack tip due to a higher TSS in the bulk region is, in a way, similar to Kim's DHC model. Even though Ambler assumed a different strain energy of the matrix hydrides with the direction of an approach to the test temperature, the peak temperature, hydrogen concentration and the hydride phase, a feasible rationale for this assumption is yet to be given. In this study, a path dependence of DHC velocity of Zr-2.5Nb tubes will be investigated using Kim's DHC model where a driving force for DHC is the supersaturated hydrogen concentration between the crack tip and the bulk region. To this ends, the furnace cooled and water-quenched Zr-2.5Nb specimens were subjected to DHC tests at different test temperatures that were approached by a heating or by a cooling. Kim's DHC model predicts that the water-quenched Zr- 2.5Nb will have DHC crack growth even at temperatures above 180 .deg. C where the furnace-cooled Zr-2.5Nb will not. This experiment will provide

Delayed Hydride Cracking in Zr-2.5Nb Tubes with the Direction of An Approach to Temperature

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Im, Kyung Soo; Kim, Kang Soo; Ahn, Sang Bok; Cheong, Yong Moo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2006-07-01

One of the unique features of delayed hydride cracking (DHC) of zirconium alloys is that the DHC velocity (DHCV) of zirconium alloys strongly depends on the path to the test temperature. Ambler reported that the DHCV of Zr-2.5Nb tubes at temperatures above 180 .deg. C depended upon the direction of an approach to the test temperatures, and reported on a presence of the DHC arrest temperature or TDAT above which the DHCV decreased upon an approach to the test temperature by a heating. Ambler proposed a hydrogen transfer from the bulk to the crack tip assuming that the hydrides formed at the crack tip and in the bulk region are fully constrained and partially constrained at the crack tip, respectively. In other words, the terminal solid solubility (TSS) of hydrogen would be governed by elastic strain energy induced by the precipitating hydrides, leading to a higher TSS in the bulk region than that at the crack tip. In a sense, his assumption that the hydrogen concentration is higher in the bulk region than that at the crack tip due to a higher TSS in the bulk region is, in a way, similar to Kim's DHC model. Even though Ambler assumed a different strain energy of the matrix hydrides with the direction of an approach to the test temperature, the peak temperature, hydrogen concentration and the hydride phase, a feasible rationale for this assumption is yet to be given. In this study, a path dependence of DHC velocity of Zr-2.5Nb tubes will be investigated using Kim's DHC model where a driving force for DHC is the supersaturated hydrogen concentration between the crack tip and the bulk region. To this ends, the furnace cooled and water-quenched Zr-2.5Nb specimens were subjected to DHC tests at different test temperatures that were approached by a heating or by a cooling. Kim's DHC model predicts that the water-quenched Zr- 2.5Nb will have DHC crack growth even at temperatures above 180 .deg. C where the furnace-cooled Zr-2.5Nb will not. This experiment

Modelling of oxidation and hydriding behaviour of Zircaloy-2 pressure tubes in PHWR

International Nuclear Information System (INIS)

Sah, D.N.; Sunil Kumar; Khan, K.B.

2002-01-01

A computer model named DOCTOR (Deuteriding of Coolant Tubes during Operation of Reactor) has been developed for predicting the axial profile of oxide thickness and hydrogen (Deuterium) concentration in PHWR pressure tubes. This model is applicable to single channel or full core analysis. The main source of hydrogen is considered to be oxidation of pressure tube on the i.d. surface by high temperature coolant water. Three stages of oxidation is considered namely, pre- transition, post transition and accelerated. Oxidation rate is considered to be dependent on channel power, axial power/flux distribution, coolant temperature and pre-existing oxide thickness at the location. The kinetics parameters for oxidation model are derived from the actual measurement of oxide thickness on a number of pressure tubes examined in PIE Division. The input data required for the model are: channel power, channel power factor, axial flux distribution, coolant inlet temperature, critical oxide thickness, hydrogen pick up fraction, initial hydrogen in the material and time of operation (efpy). The model calculates the oxide layer thickness on the inside surface of the pressure tube along the length. The amount of hydrogen picked up by the pressure tube is calculated from the oxide thickness using hydrogen pick up fraction determined from the PIE data. The pressure tube length is divided into a number of axial segments for calculation. The temperature and fast neutron flux assumed to be constant in a given segment. The axial temperature profile calculated from the axial power profile in the channel is used for calculating the oxidation rate at various locations in the pressure tube. The model has been validated with PIE data of hydrogen equivalent measurement on a number of irradiated Zircaloy-2 pressure tubes of various PHWRs. The performance of the model in predicting the axial profile of hydrogen in the pressure tubes has been found to be good. (author)

Microstructural characterization and mechanical properties of Excel alloy pressure tube material

Science.gov (United States)

Sattari, Mohammad

Microstructural characterization and mechanical properties of Excel (Zr-3.5%Sn-0.8%Mo-0.8%Nb), a dual phase alphaZr -hcp and betaZr-bcc pressure tube material, is discussed in the current study which is presented in manuscript format. Chapter 3 discusses phase transformation temperatures using different techniques such as quantitative metallography, differential scanning calorimetry (DSC), and electrical resistivity. It was found that the alphaZr → alphaZr+beta Zr and alphaZr+betaZr → betaZr transformation temperatures are in the range of 600-690°C and 960-970°C respectively. Also it was observed that upon quenching from temperatures below ˜860°C the martensitic transformation of betaZr to alpha'--hcp is halted and instead the microstructure transforms into retained Zr with o hexagonal precipitates inside betaZr grains. Chapter 4 deals with aging response of Excel alloy. Precipitation hardening was observed in samples water-quenched from high in the alphaZr+beta Zr or betaZr regions followed by aging. The optimum aging conditions were found to be 450°C for 1 hour. Transmission electron microscopy (TEM) showed dispersion of fine precipitates (˜10nm) inside the martensitic phase. Energy dispersive X-ray spectroscopy (EDS) showed the chemical composition of precipitates to be Zr-30wt%Mo-25wt%Nb-2wt%Fe. Electron crystallography using whole pattern symmetry of the convergent beam electron diffraction (CBED) patterns together with selected area diffraction (SAD) polycrystalline ring patterns, suggests the -6m2 point group for the precipitates belonging to hexagonal crystal structure, with a= 2.936 A and c=4.481 A, i.e. c/a =1.526. Crystallographic texture and high temperature tensile properties as well as creep-rupture properties of different microstructures are discussed in Chapter 5. Texture analysis showed that solution treatment high in the alpha Zr+betaZr or betaZr regions followed by water quenching or air cooling results in a more random texture compared

Evaluation of Fatigue Crack Initiation for Volumetric Flaw in Pressure Tube

International Nuclear Information System (INIS)

Choi, Sung Nam; Yoo, Hyun Joo

2005-01-01

CAN/CSA.N285.4-94 requires the periodic inservice inspection and surveillance of pressure tubes in operating CANDU nuclear power reactors. If the inspection results reveal a flaw exceeding the acceptance criteria of the Code, the flaw must be evaluated to determine if the pressure is acceptable for continued service. Currently, the flaw evaluation methodology and acceptance criteria specified in CSA-N285.05-2005, 'Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors'. The Code is applicable to zirconium alloy pressure tubes. The evaluation methodology for a crack-like flaw is similar to that of ASME B and PV Sec. XI, 'Inservice Inspection of Nuclear Power Plant Components'. However, the evaluation methodology for a blunt volumetric flaw is described in CSA-N285.05-2005 code. The object of this paper is to address the fatigue crack initiation evaluation for the blunt volumetric flaw as it applies to the pressure tube at Wolsong NPP

Self-diffusion and heterodiffusion in Zr-2.5%Nb α/β interfaces comparison with grain boundary diffusion in α-Zr

International Nuclear Information System (INIS)

Iribarren, M.J.; Dyment, F.

1991-01-01

Conventional radioactive tracer section techniques were used to make an experimental determination of diffusion parameters for Zr, Nb and Ni along the α/β boundary interfaces of Zr-2.5%Nb and comparing them with those for Zr, Nb and Co in α-Zr grain boundaries. Both determinations were made for a wide range of temperatures, including reactor working temperatures. Different materials were used for this purpose, both specially prepared alloys for diffusion experiments and part of the material from the actual pressure tubes. Different stabilizing thermal treatments were performed and results were analyzed based on the different morphologies obtained. (Author) [es

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Pressure heat pumping in the orifice pulse-tube refrigerator

International Nuclear Information System (INIS)

Boer, P.C.T. de

1996-01-01

The mechanism by which heat is pumped as a result of pressure changes in an orifice pulse-tube refrigerator (OPTR) is analyzed thermodynamically. The thermodynamic cycle considered consists of four steps: (1) the pressure is increased by a factor π 1 due to motion of a piston in the heat exchanger at the warm end of the regenerator; (2) the pressure is decreased by a factor π 2 due to leakage out of the orifice; (3) the pressure is further decreased due to motion of the piston back to its original position; (4) the pressure is increased to its value at the start of the cycle due to leakage through the orifice back into the pulse tube. The regenerator and the heat exchangers are taken to be perfect. The pressure is assumed to be uniform during the entire cycle. The temperature profiles of the gas in the pulse tube after each step are derived analytically. Knowledge of the temperature at which gas enters the cold heat exchanger during steps 3 and 4 provides the heat removed per cycle from this exchanger. Knowledge of the pressure as a function of piston position provides the work done per cycle by the piston. The pressure heat pumping mechanism considered is effective only in the presence of a regenerator. Detailed results are presented for the heat removed per cycle, for the coefficient of performance, and for the refrigeration efficiency as a function of the compression ratio π 1 and the expansion ratio π 2 . Results are also given for the influence on performance of the ratio of specific heats. The results obtained are compared with corresponding results for the basic pulse-tube refrigerator (BPTR) operating by surface heat pumping

Middle Ear Pressure Regulation - Complementary Action of the Mastoid and Eustachian Tube

DEFF Research Database (Denmark)

Gaihede, Michael; Dirckx, Joris J J; Jacobsen, Henrik

2010-01-01

, MEP counter-regulation presented as Eustachian tube openings with steep and fast pressure changes toward 0 Pa, whereas in others, gradual and slow pressure changes presented related to the mastoid; these changes sometimes crossed 0 Pa into opposite pressures. In many cases, combinations...... to continuous regulation of smaller pressures, whereas the tube was related to intermittent regulation of higher pressures....

Middle Ear Pressure Regulation - Complementary Action of the Mastoid and Eustachian Tube

DEFF Research Database (Denmark)

Gaihede, Michael; Jacobsen, Henrik; Tveterås, Kjell

, MEP counter-regulation presented as Eustachian tube openings with steep and fast pressure changes toward 0 Pa, whereas in others, gradual and slow pressure changes presented related to the mastoid; these changes sometimes crossed 0 Pa into opposite pressures. In many cases, combinations...... to continuous regulation of smaller pressures, whereas the tube was related to intermittent regulation of higher pressures....

The examination of the ruptured Zircaloy-2 pressure tube from Pickering NGS Unit 2

International Nuclear Information System (INIS)

Cheadle, B.A.; Smith, A.D.; Baskin, C.C.

1985-07-01

On 1983 August 01 a Zircaloy-2 pressure tube in Pickering NGS Unit 2 ruptured. All the fuel channel components, the fuel bundles, pressure tube, end fittings, garter springs and calandria tubes were shipped to Chalk River Nuclear Laboratories for examination to determine the cause of the rupture. The examination showed that the rupture initiated at a series of hydride blisters on the outside surface of the pressure tube. The blisters formed because of the garter spring spacers between the pressure tube and calandria tube was about one metre out of position. This allowed the horizontal pressure tube to sag by creep and touch the cool calandria tube. The resulting thermal gradients in the pressure tube concentrated the hydrogen and deuterium at the cool zones and blisters of solid hydride formed. Cracks initiated at several of the blisters and linked together to form a partial through wall critical crack which initiated the final rupture. The video presentation shows how the examination of the fuel channel components was conducted in underwater bays and shielded cells and explains the sequence of events that caused the rupture

On random pressure pulses in the turbine draft tube

Science.gov (United States)

Kuibin, P. A.; Shtork, S. I.; Skripkin, S. G.; Tsoy, M. A.

2017-04-01

The flow in the conical part of the hydroturbine draft tube undergoes various instabilities due to deceleration and flow swirling at off-design operation points. In particular, the precessing vortex rope develops at part-load regimes in the draft tube. This rope induces periodical low-frequency pressure oscillations in the draft tube. Interaction of rotational (asynchronous) mode of disturbances with the elbow can bring to strong oscillations in the whole hydrodynamical system. Recent researches on flow structure in the discharge cone in a regime of free runner had revealed that helical-like vortex rope can be unstable itself. Some coils of helix close to each other and reconnection appears with generation of a vortex ring. The vortex ring moves toward the draft tube wall and downstream. The present research is focused on interaction of vortex ring with wall and generation of pressure pulses.

Microstructural examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

Srivastava, D.; Tewari, R.; Dey, G.K.; Sah, D.N.; Banerjee, S.

2005-01-01

Irradiation induced microstructural changes in Zr alloys strongly influence the creep, growth and mechanical properties of pressure tube material. Since dimensional changes and mechanical property degradation can limit the life of pressure tube, it is essential to study and develop an understanding of the microstructure produced by neutron irradiation, by examining samples taken from the irradiated components. In the present work, an effort has been made to examine, microstructure of the Zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water Reactor (PHWR). The present work is a first step towards a comprehensive program of characterization of microstructure of reactor materials after irradiation to different fluence levels in power reactors. In this study, samples from a Zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit 1, for a period for 6.77 effective full power years (EFPYs), have been prepared and examined. The samples selected from the tube are expected to have a cumulative radiation damage of about 3 dpa. Samples prepared from the off cuts of RAPS-1 pressure tubes were also studied for examining the unirradiated microstructure of the material. The samples were examined in a 200kV JEOL 2000 FX microscope. This paper presents the distinct features observed in irradiated sample and a comprehensive comparison of the microstructures of the unirradiated and irradiated material. The effect of annealing on the annihilation of the defects generated during irradiation has been also studied. The bright field micrographs revealed that microstructure of the irradiated samples was different in many respects from the microstructure of the unirradiated samples. The presence of defect structure in the form of loops etc could be seen in the irradiated sample. These loops were mostly c-type loops lying in the basal plane. The dissolution and redistribution of the precipitates were

Measurement and analysis of pressure tube elongation in the Douglas Point reactor

International Nuclear Information System (INIS)

Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

1980-02-01

Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

Large-scale synthesis of NbS2 nanosheets with controlled orientation on graphene by ambient pressure CVD.

Science.gov (United States)

Ge, Wanyin; Kawahara, Kenji; Tsuji, Masaharu; Ago, Hiroki

2013-07-07

We report ambient pressure chemical vapor deposition (CVD) growth of single-crystalline NbS2 nanosheets with controlled orientation. On Si and SiO2 substrates, NbS2 nanosheets grow almost perpendicular to the substrate surface. However, when we apply transferred CVD graphene on SiO2 as a substrate, NbS2 sheets grow laterally lying on the graphene. The NbS2 sheets show the triangular and hexagonal shapes with a thickness of about 20-200 nm and several micrometres in the lateral dimension. Analyses based on X-ray diffraction and Raman spectroscopy indicate that the NbS2 nanosheets are single crystalline 3R-type with a rhombohedral structure of R3m space group. Our findings on the formation of highly aligned NbS2 nanosheets on graphene give new insight into the formation mechanism of NbS2 and would contribute to the templated growth of various layered materials.

Design of experiments and equipment to test the ballooning characteristics of CANDU pressure tubes

International Nuclear Information System (INIS)

Forrest, C.F.; Stern, F.; Hart, R.G.

1992-01-01

Experiments have been planned and an apparatus has been designed to enable creep testing of end-of-life pressure tube specimens in a LOCA environment. Effects that could be studied include: annealing of irradiation damage during transient heating; effects of hydride blisters on pressure tube ballooning strains; and, effects of uniformly-distributed hydrogen content on pressure tube ballooning strains. The proposed experimental program will consist of separate effects creep tests on pressure tube sections under transient heating conditions

Tube Plugging Criteria for the High-pressure Heaters of Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyungnam; Cho, Nam-Cheoul; Lee, Kuk-hee [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In this paper, a method to establish the tube plugging criteria of BOP heat exchangers is introduced and the tube plugging criteria for the high pressure heaters of a nuclear power plant. This method relies on the similar plugging criteria used in the steam generator tubes. Power generation field urges nuclear power plants to reduce operating and maintaining costs to remain competitive. To reduce the cost by means of preventing the lowering thermal efficiency, the inspection of balance-of-plant heat exchanger, which was treated as not important work, becomes important. The tubing materials and tube thickness of heat exchangers in nuclear power plants are selected to withstand system temperature, pressure, and corrosion. But tubes have experienced leaks and failures and plugged based upon eddy current testing (ET) results. There are some problems for plugging the heat exchanger tubes since the criterion and its basis are not clearly described. For this reason, the criteria for the tube wall thickness are addressed in order to operate the heat exchangers in nuclear power plant without trouble during the cycle. The feed water heater is a kind of heat exchanger which raises the temperature of water supplied from the condenser. The heat source of high-pressure heaters is the extraction steam from the high-pressure turbine and moisture separator re-heater. If the tube wall of the heater is broken, the feed water flowing inside the tube intrudes to shell side. This forces the turbine to be stop in order to protect it. There are many codes and standards to be referred for calculating the minimum thickness of the heat exchanger tube in the designing stage. However, the codes and standards related to show the tube plugging criteria may not exist currently. A method to establish the tube plugging criteria of BOP heat exchangers is introduced and the tube plugging criteria for the high pressure heaters of Ulchin NPP No. 3 and 4. This method relies on the similar plugging

Metallographic preparation of Zr-2.5Nb pressure tube material for examination of inclusions

International Nuclear Information System (INIS)

Lockley, A.J.

1994-11-01

The traditional final polish of Zr-2.5Nb alloy comprises an attack polish that contains a 0.05 μm alumina or fly-ash slurry with dilute hydrofluoric acid. This polish preferentially etches the material adjacent to the inclusions and distorts or removes the inclusions. A final polish has been developed that uses a caustic alumina slurry to produce a chemical-mechanical polish that keeps the inclusions intact. This preparation is reproducible, suitable for automation, and retains smaller inclusions. (author). 2 refs., 5 figs

The influence of material variables on corrosion and deuterium uptake of Zr-2.5Nb alloy during irradiation

Energy Technology Data Exchange (ETDEWEB)

McDougall, G.M.; Urbanic, V.F. [Chalk River Technicians Lab., AECL, Ontario (Canada)

2002-07-01

Current CANDU 2 reactors use Zr-2.5Nb pressure tubes that are extruded at 1088 K, cold-drawn 27%, and autoclaved at 673 K for 24 h. This results in a metastable, two-phase microstructure consisting of elongated {alpha}-Zr grains surrounded by a network of {beta}-Zr filaments. To develop a mathematical model of corrosion and deuterium ingress in pressure tubes, we have considered the impact of variables including: fast neutron flux, temperature, and the as fabricated microstructure and its evolution during irradiation. Small specimens of Zr-2.5Nb are being exposed under CANDU water chemistry conditions in the Halden Boiling Water Reactor. The experiments involve fast neutron fluxes (E {>=} 1.05 MeV) of 0, 1.7, and 4.5 x 10{sup 17} n x m{sup -2} - s{sup -1}, and temperatures of 523 and 598 K. Specimens have been prepared from pressure tube materials representative of all current CANDU reactors, materials subject to thermal decomposition of the {beta}-Zr phase, and tubes extruded over a range of conditions. Results from the first three years of the Halden test program are summarized. At both 523 and 598 K, tubes made of {beta}-quenched material exhibit lower oxidation rates than those made from non-{beta}-quenched materials. In short-term out-of-flux exposures at 523 K, three non-{beta}-quenched tubes appear to show linear oxidation kinetics. Similar behavior is not observed in tests conducted out-of-flux at 598 K, or in-flux at either temperature. At 598 K, {beta}-quenched tubes exhibit significantly lower deuterium pickup rates than non-{beta}-quenched tubes. When tested at 598 K, thermally aged specimens show declining oxidation and deuterium pickup rates with increasing {beta}-Zr phase decomposition. At 523 K, the impact of thermal aging was less significant. Preliminary results from an 'extrusion variable test' suggest that tubes fabricated according to the current CANDU specification show the best corrosion resistance. (authors)

Laser plasma generation of hydrogen-free diamond-like carbon thin films on Zr-2.5Nb CANDU pressure tube materials and silicon wafers with a pulsed high-power CO2 laser

International Nuclear Information System (INIS)

Ebrahim, N.A.; Mouris, J.F.; Hoffmann, C.R.J.; Davis, R.W.

1995-06-01

We report the first experiments on the laser plasma deposition of hydrogen-free, diamond-like carbon (DLC) films on Zr-2.5Nb CANDU pressure-tube materials and silicon substrates, using the short-pulse, high-power, CO 2 laser in the High-Power Laser Laboratory at Chalk River Laboratories. The films were (AFM). The thin films show the characteristic signature of DLC films in the Raman spectra obtained using a krypton-ion (Kr + ) laser. The Vickers ultra-low-load microhardness tests show hardness of the coated surface of approximately 7000 Kg force mm -2 , which is consistent with the hardness associated with DLC films. AFM examination of the film morphology shows diamond-like crystals distributed throughout the film, with film thicknesses of up to 0.5 μm generated with 50 laser pulses. With significantly more laser pulses, it is expected that very uniform diamond-like films would be produced. These experiments suggest that it should be possible to deposit hydrogen-free, diamond-like films of relevance to nuclear reactor components with a high-power and high-repetition-rate laser facility. (author). 7 refs., 2 tabs., 15 figs

Method of detecting leakage from sealing attached to pressure tube

International Nuclear Information System (INIS)

Tomomatsu, Ken-ichi; Hayashi, Ken-ichi.

1990-01-01

The present invention provides a detection method for measuring the amount of water leaked from sealings attached to the lower end of a pressure tube. That is, the lower end of the pressure tube is sealed only by a metal sealing. A capturing vessel is placed under the pressure tube for capturing the leaked water dropping from the lower end of the pressure tube and the weight of the leaked water is measured on every capturing vessels to determine the amount of the leaked water. The leakage detection method based on the weight measurement has higher accuracy compared with a conventional volume measuring method using a water level gauge as described below. For example, if the volume of the captured water is 10cc, an error of about 0.1cc is caused by the volume measuring method using the water level gauge, whereas if 10g (10cc) weight of water is measured by using an accurate balance, error is only about 10 -4 g (10 -4 cc). Accordingly, the method of the present invention can measure at an accuracy about 1000 times as high as the conventional method. (I.S.)

CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

2010-01-01

The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

Performance evaluation of reactor operated zircaloy-2 pressure tubes of RAPS-1

International Nuclear Information System (INIS)

Chatterjee, S.; Ramadasan, E.; Balakrishnan, K.S.; Bahl, J.K.

1992-01-01

Detailed post irradiation examination was carried out on pressure tube sections from E-10, F-9 and F-10 locations of RAPS-1 after an in-reactor residence equivalent to 3.6 effective full power years. The F-10 pressure tube was studied in detail on sections obtained from one end to the other, whereas in the case of E-9 and F-9 pressure tubes only the end sections were examined. The studies carried out were visual examination, metallography, hydrogen i.e. H(D) analysis and mechanical testing at 300 C. Microstructural observations revealed uniform and random hydride/deuteride platelet distribution and absence of blisters or hydride segregation. The H(D) content in the F-10 pressure tube was found to vary in the range 6-12 ppm. The typical H(D) content in the three tubes was around 1 ppm. The H(D) pick-up evaluated from the observed oxide layer thickness was 8 ppm. Longitudinal tensile specimens fabricated from the F-10 pressure tube section and tested at 300 C exhibited increase in yield strength and tensile strength of 39% and 30% respectively. The residual uniform elongation was typically 1.8%. The observed changes in the tensile properties were found to be lower than those reported on unstressed specimens irradiated to similar neutron fluences. The observed hydrogen content and tensile properties obtained in F-10 pressure tube would not be detrimental under normal reactor operating conditions. (author). 10 refs., 4 figs., 2 tabs., 1 annexure

First Research Coordination Meeting on Prediction of Axial and Radial Creep in HWR Pressure Tubes. Presentations

International Nuclear Information System (INIS)

2013-01-01

Pressure tube deformation is a critical aging issue in operating Heavy Water Reactors (HWRs). According to the service year, horizontal pressure tubes have three kinds of deformation: diametral creep leading to the flow bypass and the penalty to critical heat flux for fuel rods, longitudinal creep leading to the interference of feeder pipes and/or with fuelling machine, and sagging leading to the interference with in-core components and potential contact between the pressure tube and calandria tube. The CRP scope includes the establishment of a database for pressure tube deformation, microstructure characterization of pressure tube materials collected from HWRs currently operating in Member States and development of a prediction model for pressure tube deformation

Total hemispherical emissivity of pre-oxidized and un-oxidized Zr-2.5Nb pressure-tube materials at 600 {sup o}C to 1000 {sup o}C under vacuum

Energy Technology Data Exchange (ETDEWEB)

Fong, R.W.L.; Paine, M.; Nitheanandan, T., E-mail: randy.fong@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

The emissivity of pre-oxidized and un-oxidized pressure-tube specimens has been measured at high temperatures under vacuum. The emissivity values of un-oxidized tube specimens decreased only slightly from 0.34 at 600 {sup o}C to 0.30 at 800 {sup o}C and changed gradually to 0.25 at 1000 {sup o}C. In comparison, the emissivity of pre-oxidized pressure-tube specimens decreased drastically from 0.70 at 600 {sup o}C to 0.35 at 800 {sup o}C, and gradually decreased to 0.25 at 1000 {sup o}C. The oxide layer of the pre-oxidized tube specimens dissolved into the metal matrix when heated to 700 {sup o}C and higher. Using these results, 2 linear correlations were obtained for emissivity with the oxide thickness measured by scanning electron microscopy and secondary ion mass spectroscopy analysis. (author)

Investigation of pressure drop in capillary tube for mixed refrigerant Joule-Thomson cryocooler

International Nuclear Information System (INIS)

Ardhapurkar, P. M.; Sridharan, Arunkumar; Atrey, M. D.

2014-01-01

A capillary tube is commonly used in small capacity refrigeration and air-conditioning systems. It is also a preferred expansion device in mixed refrigerant Joule-Thomson (MR J-T) cryocoolers, since it is inexpensive and simple in configuration. However, the flow inside a capillary tube is complex, since flashing process that occurs in case of refrigeration and air-conditioning systems is metastable. A mixture of refrigerants such as nitrogen, methane, ethane, propane and iso-butane expands below its inversion temperature in the capillary tube of MR J-T cryocooler and reaches cryogenic temperature. The mass flow rate of refrigerant mixture circulating through capillary tube depends on the pressure difference across it. There are many empirical correlations which predict pressure drop across the capillary tube. However, they have not been tested for refrigerant mixtures and for operating conditions of the cryocooler. The present paper assesses the existing empirical correlations for predicting overall pressure drop across the capillary tube for the MR J-T cryocooler. The empirical correlations refer to homogeneous as well as separated flow models. Experiments are carried out to measure the overall pressure drop across the capillary tube for the cooler. Three different compositions of refrigerant mixture are used to study the pressure drop variations. The predicted overall pressure drop across the capillary tube is compared with the experimentally obtained value. The predictions obtained using homogeneous model show better match with the experimental results compared to separated flow models

Status and Plans for work on pressure tube creep at AECL

International Nuclear Information System (INIS)

Bickel, Grant A.

2013-01-01

AECL research goals: • Develop empirical models to: – regress out operating conditions/extrinsic factors – rank relative strain behavior of measured in-service pressure tubes; • Correlate the ranked strains to manufacturing variables and the microstructure to: – Develop mechanistic insights – Optimize manufacturing/microstructure for improved pressure tube performance

Gamma sensitivity of pressurized drift tubes

International Nuclear Information System (INIS)

Baranov, S.A.; Bojko, I.R.; Shelkov, G.A.; Ignatenko, M.A.

1995-01-01

Using a set of commonly used radioactive sources, the efficiency of pressurized drift tubes for gammas with energy from 5.9 keV up to 1.3 MeV has been measured. The tube was made of aluminium and filled with Ar, 15%CO 2 and 2.5%iC 4 H 10 gas mixture at 3 atm. The measured efficiency is compared with the results of the calculations in the frame of our simple model as well as with that of the Monte Carlo simulation using GEANT code. The results of our calculations are in agreement with experimental data, while GEANT simulation tends to give lower efficiency in the energy range of 200 keV γ <1300 keV. The average efficiency of the tube in the field of ATLAS gamma background is about 0.45%. 8 refs., 7 figs., 1 tab

Refrigerant charge, pressure drop, and condensation heat transfer in flattened tubes

Energy Technology Data Exchange (ETDEWEB)

Wilson, M J; Newell, T A; Chato, J C [University of Illinois, Urbana, IL (United States). Dept. of Mechanical and Industrial Engineering; Infante Ferreira, C A [Delft University of Technology (Netherlands). Laboratory for Refrigeration and Indoor Climate Control

2003-06-01

Horizontal smooth and microfinned copper tubes with an approximate diameter of 9 mm were successively flattened in order to determine changes in flow field characteristics as a round tube is altered into a flattened tube profile. Refrigerants R134a and R410A were investigated over a mass flux range from 75 to 400 kg m{sup -2} s{sup -}2{sup 1} and a quality range from approximately 10-80%. For a given refrigerant mass flow rate, the results show that a significant reduction in refrigerant charge is possible. Pressure drop results show increases of pressure drop at a given mass flux and quality as a tube profile is flattened. Heat transfer results indicate enhancement of the condensation heat transfer coefficient as a tube is flattened. Flattened tubes with an 18{sup o} helix angle displayed the highest heat transfer coefficients. Smooth tubes and axial microfin tubes displayed similar levels of heat transfer enhancement. Heat transfer enhancement is dependent on the mass flux, quality and tube profile. (author)

Pressure relief experiments on a cyclindrical carbon brick tube

International Nuclear Information System (INIS)

Lang, H.; Weise, H.J.; Ennen, P.

1978-08-01

Pressure relief experiments have been carried out on a carbon brick tube. The outer diameter of the specimen was 580 mm, the inner diameter 280 mm, the length 800 mm. The experiments were made with helium at the temperature of the environment. The measurements were carried out in the pressure range from 15 upto 39 bar. The pressure loss was measured dependent on the initial pressure and on time at 5 positions uniformly distributed over the thickness of the tube wall and in the pressure vessel. The maximum pressure transients occurred amounted to approximately 60 bar/second. The maximum overpressure with respect to the environment which occurred in the carbon brick during the relief experiments was about 3.3 bar. The measurements distinctly showed the presence and the effects of inhomogeneities in the sample material. No damages or changes in the carbon brick, which could be regarded as a consequence of the experiments, were found. (orig./GSC) [de

[Characterizing the passive opening of the eustachian tube in a hypo-/hyperbaric pressure chamber].

Science.gov (United States)

Meyer, M F; Mikolajczak, S; Luers, J C; Lotfipour, S; Beutner, D; Jumah, M D

2013-09-01

Beside arbitrary and not arbitrary active pressure equalization systems there is a passive equalization system via the Eustachian tube (ET) at pressure difference between the epipharyngeal space and the middle ear. Aim of this study was to characterize this passive equalization system in a hypobaric/hyperbaric pressure chamber by continuously measuring the tympanic impedance. In contrast to other studies, which are measured only in a hypobaric pressure chamber it is possible to include participants with Eustachian tube dysfunction (ETD). Following a fixed pressure profile 39 participants were exposed to phases of pressure rising and decompression. By continuously measuring the tympanic impedance in the pressure chamber it was possible to measure data of the Eustachian Tube opening Pressure (ETOP), Eustachian Tube closing pressure (ETCP) and Eustachian Tube opening duration (ETOD). In addition it was possible to characterize the gradient of pressure during decompression, while the ET was open. Beside the measurement of the arithmetic average of the ETOP (30.2 ± 15.1 mbar), ETCP (9.1 ± 7.7 mbar) and ETOD (0.65 ± 0.38 s) it was obvious that there are recurrent samples of pressure progression during the phase of tube opening. Generally it is possible to differentiate between the type of complete opening and partial opening. The fundamental characterization of the action of the passive tube opening, including the measurement of the ETOP, ETCP and ETOD, is a first step in understanding the physiological and pathophysiological function of the ET. © Georg Thieme Verlag KG Stuttgart · New York.

Pressure tube replacement in Pickering NGS A units 1 and 2

International Nuclear Information System (INIS)

Irvine, H.S.; Bennett, E.J.; Talbot, K.H.

1986-10-01

Being able to technically and economically replace the most radioactive components (excluding the nuclear fuel) in operating reactors will help to ensure the ongoing acceptance of nuclear power as a viable energy source for the future. Ontario Hydro is well along the path to meeting the above objective for its CANDU-PHW reactors. Following the failure of a Zircaloy-II pressure tube in unit 2 of Pickering NGS A in August, 1983, Ontario Hydro has embarked on a program to replace all Zircaloy-II pressure tubes in units 1 and 2 at Pickering. This program integrates the in-house research, design, construction, and operating skills of a large utility (Ontario Hydro) with the skills of a national nuclear organization (Atomic Energy of Canada Limited) and the private engineering sector of the Canadian nuclear industry. The paper describes the background to the pressure tube failure in Pickering unit 2 and to the efforts incurred in understanding the failure mechanism and how similar failures are not expected for the zirconium-niobium pressure tube material used in all other large CANDU-PHW units after units 1 and 2 of Pickering NGS A. The tooling developed for the pressure tube replacement program is described as well as the organization to undertake the program in an operating nuclear station. The retubing of units 1 and 2 at Pickering NGS A is nearing a successful completion and shows the benefits of being able to integrate the various skills required for this success. Pressure tube replacement in a CANDU-PHW reactor is equivalent to replacement of the reactor vessel in a LWR. The fact that this replacement can be done economically and with acceptable radiation dose to workers augurs well for the continued viability of the use of nuclear energy for the benefit of mankind. (author)

Decontamination and recycle of zirconium pressure tubes from Pressurized Heavy Water Reactor

International Nuclear Information System (INIS)

Gantayet, L.M.; Verma, R.; Remya Devi, P.S.; Banerjee, S.; Kotak, V.; Raha, A.; Sandeep, K.C.; Joshi, Shreeram W.; Lali, A.M.

2009-01-01

An ion exchange process has been developed for decontamination of zirconium pressure tubes from Pressurized Heavy Water Reactor and recycling of neutronically improved zirconium. Distribution coefficient, equilibrium isotherm, kinetic and breakthrough data were used to develop the separation process. Effect of gamma radiation on indigenous resins was also studied to assess their suitability in high radiation field. (author)

Flooding of a large, passive, pressure-tube LWR

Energy Technology Data Exchange (ETDEWEB)

Hejzlar, P.; Todreas, N.E.; Driscoll, M.J. [Massachusetts Institute of Technology, Cambridge, MA (United States)

1995-09-01

A reactor concept has been developed which can survive LOCA without scram and without replenishing primary coolant inventory. The proposed concept is a pressure tube type reactor similar to CANDU reactors, but differing in three key aspects: (1) a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles, (2) the heavy water coolant in the pressure tubes is replaced by light water, and (3) the calandria tank contains a low pressure gas instead of heavy water moderator. The gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents, it allows passive calandria flooding. This paper describes the thermal hydraulic characteristics of the gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube LWR concept. The flooding of the top row of fuel channels must be accomplished fast enough so that none of the critical components of the fuel channel exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. Two other considerations are important. The thermal shock experienced by the calandria and pressure tubes has been evaluated and shown to be within acceptable bounds. Finally, although complete flooding renders the reactor deeply subcritical, various steam/water densities can be hypothesized to be present during the flooding process which could cause reactivity to increase from the initially voided calandria case. One such hypothesis which leads to the maximum possible density of the steam/water mixture in the still unflooded calandria space is entrainment from the free surface. It is shown that the steam/water mixture density yielding the maximum reactivity peak cannot be achieved by entrainment because it exceeds thermohydraulically attainable densities of steam/water by an order of magnitude.

Impulse Pressuring Diffusion Bonding of TiC Cermet to Stainless Steel Using Ti/Nb Interlayer

Directory of Open Access Journals (Sweden)

LI Jia

2017-03-01

Full Text Available Impulse pressuring diffusion bonding(IPDB and constant pressuring diffusion bonding(CPDB of TiC cermet to 304 stainless steel(304SS using Ti/Nb interlayer was carried out at 890℃ under a impulse pressuring of 2-10MPa and a constant pressuring of 10MPa within a duration of only 4-12min, and a robust metallurgical bonding was achieved. Microstructure characterization and shear performance of the IPDB and CPDB joints were analyzed by SEM, EDS, XRD and shearing test. The results show that the interface phases in those two kinds of joints are similar, which are mainly σ phase,(β-Ti, Nb and α+β-Ti solid solution. When the joint is bonded for 10min, shear strength of TiC/304SS CPDB joints is 55.6MPa, while the shear strength of IPDB joints reaches 110MPa. The fracture of CPDB joints is TiC cermet fracture, while that of IPDB joints is mixed fracture by alternated between TiC cermet and reaction layer.

Calandria cooling structure in pressure tube reactor

International Nuclear Information System (INIS)

Hyugaji, Takenori; Sasada, Yasuhiro.

1976-01-01

Purpose: To contrive the structure of a heavy water distributing device in a pressure tube reactor thereby to reduce the variation in the cooling function thereof due to the welding deformation and installation error. Constitution: A heating water distributing plate is provided at the lower part of the upper tubular plate of a calandria tank to form a heavy water distributing chamber between both plates and a plurality of calandria tubes. Heavy water which has flowed in the upper part of the heavy water distributing plate from the heavy water inlet nozzle flows down through gaps formed around the calandria tubes, whereby the cooling of the calandria tank and the calandria tubes is carried out. In the above described calandria cooling structure, a heavy water distributing plate support is provided to secure the heavy water distributing plate and torus-shaped heavy water distributing rings are fixed to holes formed in the heavy water distributing plate penetrating through the calandria tubes thereby to form torus-shaped heavy water outlet ports each having a space. (Seki, T.)

Moderator mixing after a pressure tube failure

International Nuclear Information System (INIS)

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

1997-01-01

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

Leakage Characteristics of Dual-Cannula Fenestrated Tracheostomy Tubes during Positive Pressure Ventilation: A Bench Study

Directory of Open Access Journals (Sweden)

Thomas Berlet

2016-01-01

Full Text Available This study compared the leakage characteristics of different types of dual-cannula fenestrated tracheostomy tubes during positive pressure ventilation. Fenestrated Portex® Blue Line Ultra®, TRACOE® twist, or Rüsch® Traceofix® tracheostomy tubes equipped with nonfenestrated inner cannulas were tested in a tracheostomy-lung simulator. Transfenestration pressures and transfenestration leakage rates were measured during positive pressure ventilation. The impact of different ventilation modes, airway pressures, temperatures, and simulated static lung compliance settings on leakage characteristics was assessed. We observed substantial differences in transfenestration pressures and transfenestration leakage rates. The leakage rates of the best performing tubes were <3.5% of the delivered minute volume. At body temperature, the leakage rates of these tracheostomy tubes were <1%. The tracheal tube design was the main factor that determined the leakage characteristics. Careful tracheostomy tube selection permits the use of fenestrated tracheostomy tubes in patients receiving positive pressure ventilation immediately after stoma formation and minimises the risk of complications caused by transfenestration gas leakage, for example, subcutaneous emphysema.

Study of creep collapse of tubes subject to external pressure at elevated temperature

International Nuclear Information System (INIS)

Takikawa, N.

1982-01-01

Intermediate heat exchanger (IHX) tubes of VHTR form the boundary between the primary and secondary coolants of the reactor. The tubes are subject to external pressures at a postulated secondary coolant depressurization accident, which might lead to creep collapse. Therefore, it is necessary to ensure the integrity against creep collapse by analysis. The objective of this work is to study a simplified analytical method for predicting collapse time of a curved tube subjected to an external pressure. The study is made based on the comparison of experimental collapse time of curved and straight tubes. Creep collapse tests were conducted under an elevated temperature and an external pressure. Test results showed that curved tubes had longer collapse time than straight tubes with the same cross sectional ovality. The simplified analytical method for a curved tube is proposed in this report, which is to compute collapse time of a straight tube with the same ovality. And in this method the computed time is considered as collapse time of the curved tube. The above test results show that this simplified method gives the conservative collapse time. And it is confirmed by additional IHX tube tests that the method is applicable to creep collapse analysis of IHX tubes

Evaluation of techniques for inspection and diagnostics of HWR pressure tubes

International Nuclear Information System (INIS)

Choi, Jong-Ho

2008-01-01

Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes for Heavy Water Reactors (HWRs), are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Project (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. The objective of the CRP was to inter-compare inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of the CRP, participants investigated the capability of different techniques to detect and characterize flaws. During the second phase, participants collaborated to determine the hydrogen concentration and to detect and characterize hydride blisters in zirconium alloy pressure tubes. Eight organizations from six countries, which operate HWRs, have participated in this CRP, Most of the techniques examined are well established and many of them are regularly used during in-service inspection of pressure tubes. The inter-comparison of these techniques provides a platform for identifying a particular technique (or a set of techniques), which is more accurate and reliable as compared to others for a specified task. The CRP also witnessed some new methodologies, which can be implemented on in-service inspection tools. These new techniques could complement the existing ones to overcome their limitations, thereby improving the reliability and accuracy of in-service inspection. This CRP also identified future areas of research and development. (author)

Low-field Instabilities in Nb$_{3}$Sn Multifilamentary Wires the Possible Role of Unreacted Nb

CERN Document Server

Devred, A; Celentano, G; Fabbricatore, P; Ferdeghini, C; Greco, M; Gambardella, U

2007-01-01

We report an experimental study aiming to demonstrate the not negligible role of unreacted Nb on the magnetic instabilities in superconducting Nb$_{3}$Sn multifilamentary wires, observable through partial flux jumps at magnetic field values below 0.5 T. The analysed wires were recently developed for use as dipoles required in future high-energy proton accelerators and are based on powder-in-tube technology. We studied both unreacted (only involving Nb filaments) and reacted wires, finding flux jump instabilities in both cases when performing magnetic measurements. The results can be interpreted on the basis of the critical state model and are coherent with the intrinsic stability criterion.

Deadly pressure pneumothorax after withdrawal of misplaced feeding tube

DEFF Research Database (Denmark)

Andresen, Erik Nygaard; Frydland, Martin; Usinger, Lotte

2016-01-01

BACKGROUND: Many patients have a nasogastric feeding tube inserted during admission; however, misplacement is not uncommon. In this case report we present, to the best of our knowledge, the first documented fatality from pressure pneumothorax following nasogastric tube withdrawal. CASE PRESENTATION......, but our patient died less than an hour after withdrawal. The autopsy report stated that cause of death was tension pneumothorax, which developed following withdrawal of the misplaced feeding tube. CONCLUSIONS: The indications for insertion of nasogastric feeding tubes are many and the procedure...... is considered harmless; however, if the tube is misplaced there is good reason to be cautious on removal as this can unmask puncture of the pleura eliciting pneumothorax and, as this case report shows, result in an ultimately deadly tension pneumothorax....

Conservatism in methodologies for moderator subcooling sufficiency for fuel channel integrity upon pressure tube and calandria tube contact

Energy Technology Data Exchange (ETDEWEB)

Sun, L., E-mail: LSun@nbpower.com [Point Lepreau Generating Station, Lepreau, NB, (Canada)

2015-07-01

During a postulated large LOCA event in CANDU reactors, the pressure tube may balloon to contact with its surrounding calandria tube to transfer heat to the moderator. To confirm the integrity of the fuel channel in this case, many experiments have been performed in the last three decades. Based on the extant database of the pressure tube/calandria tube (PT/CT) contact, an analytical methodology was developed by Canadian Nuclear Industry to determine the sufficiency of moderator subcooling for fuel channel integrity. At the same time a semi-empirical methodology with an idea of Equivalent Moderator Subcooling (EMS) was also developed to judge the sufficiency of the moderator. In this work, some discussions were made over the two methodologies on their conservatism and it is demonstrated that the analytical approach is over conservative comparing with the EMS methodology. By using the EMS methodology, it is demonstrated that applying glass-peened calandria tubes, the requirement to moderator subcooling can be reduced by 10{sup o}C from that for smooth calandria tubes. (author)

Holographic NDE of pressure tubes for Cirene nuclear reactor

International Nuclear Information System (INIS)

Di Chirico, G.; Pirodda, L.; Villani, A.

1985-01-01

Pressure tubes for CIRENE nuclear reactor can be subjected to fretting corrosion of the inner walls. The resulting marks exhibit different geometries, whose influence on the structural behaviour of the tubes has been evaluated by means of a real time holographic technique. The paper shows the results of this investigation. Position and shape of internal defects have been directly visualized by observing holographic fringe distorsions on the outside surface of the tubes. Furthermore, through the fringe patterns, circumferential stress values have also been obtained. (Author) [pt

Fabrication of Nb3Al superconducting wires by utilizing the mechanically alloyed Nb(Al)ss supersaturated solid-solution with low-temperature annealing

International Nuclear Information System (INIS)

Pan, X.F.; Yan, G.; Qi, M.; Cui, L.J.; Chen, Y.L.; Zhao, Y.; Li, C.S.; Liu, X.H.; Feng, Y.; Zhang, P.X.; Liu, H.J.

2014-01-01

Highlights: • This paper reported superconducting properties of the powder-in-tube Nb 3 Al wires. • The Nb 3 Al wires were made by using Nb(Al) ss supersaturated solid solution powders. • The Cu-matrix Nb 3 Al superconducting wires have been successfully fabricated. • The transport J c of Nb 3 Al wires at 4.2 K, 10 T is up to 12,700 A/cm 2 . - Abstract: High-performance Nb 3 Al superconducting wire is a promising candidate to the application of high-field magnets. However, due to the production problem of km-grade wires that are free from low magnetic field instability, the Nb 3 Al wires made by rapid heating, quenching and transformation (RHQT) are still not available to the large-scale engineering application. In this paper, we reported the properties of the in situ powder-in-tube (PIT) Nb 3 Al superconducting wires, which were made by using the mechanically alloyed Nb(Al) ss supersaturated solid solution, as well as the low temperature heat-treatment at 800 °C for 10 h. The results show that Nb 3 Al superconductors in this method possess very fine grains and well superconducting properties, though a little of Nb 2 Al and Nb impurities still keep being existence at present work. At the Nb 3 Al with a nominal 26 at.% Al content, the onset T c reaches 15.8 K. Furthermore, a series of Nb 3 Al wires and tapes with various sizes have been fabricated; for the 1.0 mm-diameter wire, the J c at 4.2 K, 10 T and 14 T have achieved 12,700 and 6900 A/cm 2 , respectively. This work suggests it is possible to develop high-performance Cu-matrix Nb 3 Al superconducting wires by directly using the Nb(Al) ss supersaturated solid-solution without the complex RHQT heat-treatment process

Boussignac continuous positive airway pressure for weaning with tracheostomy tubes

NARCIS (Netherlands)

Dieperink, Willem; Aarts, Leon P. H. J.; Rodgers, Michael G. G.; Delwig, Hans; Nijsten, Maarten W. N.

2008-01-01

Background: In patients who are weaned with a tracheostomy tube ( TT), continuous positive airway pressure ( CPAP) is frequently used. Dedicated CPAP systems or ventilators with bulky tubing are usually applied. However, CPAP can also be effective without a ventilator by the disposable Bous-signac

Pressure effects on the CDW transitions and magnetoresistances in NbSe3

International Nuclear Information System (INIS)

Yasuzuka, S.; Okajima, Y.; Tanda, S.; Yamaya, K.

1999-01-01

Magnetoresistance (MR) of NbSe 3 has been measured between 2 K and 150 K under high pressure including critical pressures (P c2 and P c1 ) where the lower and upper CDW phase are totally suppressed by pressure, respectively. At ambient pressure, a large MR (LMR) is observed just below T c2 , but no MR is detected above T c2 . Under high pressure, in addition to the LMR, we observe newly a MR above T c2 . Here we call it the pressure-induced MR (PIMR). The LMR always appears as long as the lower CDW phase exists, but it rapidly disappears above P c2 . Similarly, the PIMR disappears above P c1 . Furthermore, no anomalies associated with the field-induced CDW are observed within our experimental limits. From these findings, we claim that the LMR and PIMR are due to normal carriers in small pockets created by a pressure-dependent imperfect nesting of the Fermi surface in the CDW transitions. (orig.)

Propagation of atmospheric-pressure ionization waves along the tapered tube

Science.gov (United States)

Xia, Yang; Wang, Wenchun; Liu, Dongping; Yan, Wen; Bi, Zhenhua; Ji, Longfei; Niu, Jinhai; Zhao, Yao

2018-02-01

Gas discharge in a small radius dielectric tube may result in atmospheric pressure plasma jets with high energy and density of electrons. In this study, the atmospheric pressure ionization waves (IWs) were generated inside a tapered tube. The propagation behaviors of IWs inside the tube were studied by using a spatially and temporally resolved optical detection system. Our measurements show that both the intensity and velocity of the IWs decrease dramatically when they propagate to the tapered region. After the taper, the velocity, intensity, and electron density of the IWs are improved with the tube inner diameter decreasing from 4.0 to 0.5 mm. Our analysis indicates that the local gas conductivity and surface charges may play a role in the propagation of the IWs under such a geometrical constraint, and the difference in the dynamics of the IWs after the taper can be related to the restriction in the size of IWs.

Thermal behaviour of pressure tube under fully and partially voided heating conditions using 19 pin fuel element simulator

International Nuclear Information System (INIS)

Yadav, Ashwini K.; Kumar, Ravi; Gupta, Akhilesh; Chatterjee, B.; Mukhopadhya, D.; Lele, H.G.

2011-01-01

In a nuclear reactor temperature can rise drastically during LOCA due to failure of heat transportation system and subsequently leads to mechanical deformations like sagging, ballooning and breaching of pressure tube. To understand the phenomenon an experiment has been carried out using 19 pin fuel element simulator. Main purpose of the experiment was to trace temperature profiles over the pressure tube, calandria tube and clad tubes of 220 MWe Indian Pressurised Heavy Water Reactor (IPHWR). The symmetrical heating of pressure tube of 1 m length was done through resistance heating of 19 pins under 13.5 kW power using a rectifier and the variation of temperatures over the circumference of pressure tube (PT), calandria tube (CT) and clad tubes were measured. The sagging of pressure tube was initiated at 460 deg C temperature and highest temperature attained was 650 deg C. The highest temperature attained by clad tubes was 680 deg C (over outer ring) and heat is dissipated to calandria vessel mainly due to radiation and natural convection. Again to simulate partially voided conditions, asymmetrical heating of pressure was carried out by injecting 8 kW power to upper 8 pins of fuel simulator. A maximum temperature difference of 295 deg C was observed over the circumference of pressure tube which highlights the magnitude of thermal stresses and its role in breaching of pressure tube under partially voided conditions. Integrity of pressure tube was retained during both symmetrical and asymmetrical heatup conditions. (author)

Stress relaxation analysis and irradiation creep and swelling in pressure tubes

International Nuclear Information System (INIS)

Beeston, J.M.; Burr, T.K.

1979-01-01

An analysis is presented of slit width test information on two pressure tubes that had been irradiated in test reactors. The analysis showed that differential swelling stresses and thermal stresses undergo relaxation. The mechanism responsible for the stress relaxation at temperatures less than 700 K was irradiation creep. Irradiation creep in thermal test reactor pressure tubes is evidently greater than it would be at equivalent conditions in fast reactors. The residual stresses observed in the slit width tests varied between 30 and 257 MPa and would act to reduce the operating stresses, thus allowing for increased service life of the tubes as compared with no stress relaxation

Garter spring location of pressure tube for PHWR using eddy current testing methods

International Nuclear Information System (INIS)

Lee, Y. S.; Yang, D. J.; Jeong, H. K.

2001-01-01

There are garter springs between pressure tube and calandria tube for PHWR. If the space of these garter springs become to be changed, the sagging of tube is caused and the contact between the pressure tube and calandria tube will cause the tube to be failed. AECL has applied the eddy current testing methods using send-receive type probe for this purpose, but this study apply eddy current testing methods using bobbin differential type probe to detection of garter spring location. And we did the computer simulation using VIC-3D code and compared it with experiments results for inspection 1 ∼ 11kHz. The results was that the garter spring signal was successfully detected for every frequency, and 5 kHz was best

Neutron imaging of Zr-1%Nb fuel cladding material containing hydrogen

International Nuclear Information System (INIS)

Svab, E.; Meszaros, Gy.; Somogyvari, Z.; Balasko, M.; Koeroesi, F.

2004-01-01

Hydrogen distribution and hydride phases were analyzed in reactor fuel cladding pressure tube Zr-1%Nb material up to 13,300 ppm. From neutron diffraction measurements, formation of cubic δ-ZrH 2 and a small amount of tetragonal γ-ZrH was established. Texture effects were analyzed by imaging plate technique. From neutron radiography images a linear model was set up that adequately described the relationship between gray levels and nominal H-concentrations. The H-distribution was unveiled by 3D intensity histograms and fractal analysis of multilevel-segmented neutron radiography images

Shock tubes: compressions in the low pressure chamber

International Nuclear Information System (INIS)

Schins, H.; Giuliani, S.

1986-01-01

The gas shock tube used in these experiments consists of a low pressure chamber and a high pressure chamber, divided by a metal-diaphragm-to-rupture. In contrast to the shock mode of operation, where incident and reflected shocks in the low pressure chamber are studied which occur within 3.5 ms, in this work the compression mode of operation was studied, whose maxima occur (in the low pressure chamber) about 9 ms after rupture. Theoretical analysis was done with the finite element computer code EURDYN-1M, where the computation was carried out to 30 ms

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

International Nuclear Information System (INIS)

Gulshani, P.; So, C.B.

1986-10-01

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

Methodologies for assessment of the service life of pressure tubes in Indian PHWRs

International Nuclear Information System (INIS)

Sinha, R.K.; Sharma, A.; Madhusoodanan, K.; Sinha, S.K.; Malshe, U.D.

1997-01-01

For estimating safe service life of pressure tubes in Indian PHWRs, analytical methodologies have been developed to evaluate creep deformation, deuterium pick-up rate, blister growth at cold spot, and operating domain required for achieving leak-before-break. The paper provides an overview of these methodologies, and results of some studies carried out towards evolution of proposed fitness-for-service criteria for a pressure tube in contact with its calandria tube. (author)

Heat transfer and pressure drop of condensation of hydrocarbons in tubes

Science.gov (United States)

Fries, Simon; Skusa, Severin; Luke, Andrea

2018-03-01

The heat transfer coefficient and pressure drop are investigated for propane. Two different mild steel plain tubes and saturation pressures are considered for varying mass flux and vapour quality. The pressure drop is compared to the Friedel-Correlation with two different approaches to determine the friction factor. The first is calculation as proposed by Friedel and the second is through single phase pressure drop investigations. For lower vapour qualities the experimental results are in better agreement with the approach of the calculated friction factor. For higher vapour qualities the experimental friction factor is more precise. The pressure drop increases for a decreasing tube diameter and saturation pressure. The circumferential temperature profile and heat transfer coefficients are shown for a constant vapour quality at varying mass fluxes. The subcooling is highest for the bottom of the tube and lowest for the top. The average subcooling as well as the circumferential deviation decreases for rising mass fluxes. The averaged heat transfer coefficients are compared to the model proposed by Thome and Cavallini. The experimental results are in good agreement with both correlations, however the trend is better described with the correlation from Thome. The experimental heat transfer coefficients are under predicted by Thome and over predicted by Cavallini.

Irradiation effects on Zr-2.5Nb in power reactors

Energy Technology Data Exchange (ETDEWEB)

Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

Detection of pressure tube leaks relying on moisture beetles only

International Nuclear Information System (INIS)

Kenchington, J.M.; Choi, A.; Jin, Y.

2004-01-01

A major decision was made for Pickering NGS A Annulus Gas System (ACS) that detection of a pressure tube (PT) leak should be achieved by using only moisture beetles and that dew point monitors would provide 'early warning' without status to shut down the reactor. Experience with Unit 3 has shown that dew point monitoring of pressure tube leaks was particularly subject to gas leaks and surface adsorption effects. Unit 4 was the first one to be converted during the full scale pressure tube replacement programme. Because of the fundamental change in design philosophy, moisture injection tests were carried out during commissioning to demonstrate that performance matched design. In particular it was necessary to show that leak before break (LBB) would be achieved if a leak occurred in the limiting string. Units 1 and 3 have since been converted. No decision has been taken to convert Pickering B units as gas leaks are small and no significant adsorption effects are anticipated. Hence dew point monitoring will not be impaired. (author)

Cu-Nb3Sn superconducting wires prepared by ''Copper Liquid Phase Sintering method'' using the Nb-H

International Nuclear Information System (INIS)

Resende, A.T. de.

1985-01-01

Cu-30% Nb in weighting were prepared by the method of Copper sintering liquid phase the method was improved by substitution of Nb power by Nb-H powder, obtaining a high density material with good mechanical properties, which was reduced to fine. Wire, Without heat treatment. The Cu-Nb 3 Sn wires were obtained by external diffusion process depositing tin in the Cu-30%Nb wires, and by internal diffusion process using the Sn-8.5% Cu in weighting, which was reduced to rods of 3.5 mm. These Cu-30%Nb rods were enclosed in copper tubes and deformed mechanically by rotary swaging and drawing. During the drawing step some wires were fractured, that were analysed and correlated with the microstructure of the Sn-8.5 Wt% Cu alloy. External and internal diffusion samples; after a fast thermal treatment for Sn diffusion, were submited to the temperature of 700 0 C to provide the reaction between Sn and Nb, leading to the Nb 3 Sn phase. Samples with several reaction times, and its influence on T c and J c critical parameters and normal resistivity were prepared and analysed. (author) [pt

Application of the VAW tube digester for metallurgical pressure-leaching processes

International Nuclear Information System (INIS)

Kaempf, F.; Pietsch, H.B.

1978-01-01

Problems associated with the treatment of complex and refractory ores or concentrates, as well as those related to environmental factors, have led to increased interest in hydrometallurgy under elevated temperatures and pressures. Pressure leaching can be carried out in vertical, horizontal or spherical autoclaves equipped with mechanical agitators. If high throughput capacities are catered for, the division of a conventional plant into several units is inevitable. By contrast, the VAW (Vereinigte Aluminium-Werke Aktiengesellschaft) tube digester enables hydrometallurgical processes to be carried out under pressure and at a high temperature with the use of a basically simple technology, extremely high specific throughput and improved thermal economics being achieved. The advantages of the tube digester over vessel autoclaves are described, and details of laboratory investigations into the applicability of tube digesters to various metallurgical applications are given. Test results are given for the leaching of refractory uranium ores. (author)

Pressure tube reactor

International Nuclear Information System (INIS)

Matsumoto, Tomoyuki; Fujino, Michihira.

1980-01-01

Purpose: To equalize heavy water flow distribution by providing a nozzle for externally injecting heavy water from a vibration preventive plate to the upper portion to feed the heavy water in a pressure tube reactor and swallowing up heavy water in a calandria tank to supply the heavy water to the reactor core above the vibration preventive plate. Constitution: A moderator injection nozzle is mounted on the inner wall of a calandria tank. Heavy water is externally injected above the vibration preventive plate, and heavy water in the calandria tank is swallowed up to supply the heavy water to the core reactor above the vibration preventive plate. Therefore, the heavy water flow distribution can be equalized over the entire reactor core, and the distribution of neutron absorber dissolved in the heavy water is equalized. (Yoshihara, H.)

Fabrication of Nb{sub 3}Al superconducting wires by utilizing the mechanically alloyed Nb(Al){sub ss} supersaturated solid-solution with low-temperature annealing

Energy Technology Data Exchange (ETDEWEB)

Pan, X.F. [National Engineering Laboratory for Superconducting Material, Western Superconducting Technologies (WST) Co., Ltd., Xi’an 710018 (China); Superconducting Materials Center, Northwest Institute for Nonferrous Metal Research, Xi’an 710016 (China); Yan, G., E-mail: gyan@c-nin.com [National Engineering Laboratory for Superconducting Material, Western Superconducting Technologies (WST) Co., Ltd., Xi’an 710018 (China); Superconducting Materials Center, Northwest Institute for Nonferrous Metal Research, Xi’an 710016 (China); Qi, M. [Superconducting Materials Center, Northwest Institute for Nonferrous Metal Research, Xi’an 710016 (China); Cui, L.J. [National Engineering Laboratory for Superconducting Material, Western Superconducting Technologies (WST) Co., Ltd., Xi’an 710018 (China); Chen, Y.L.; Zhao, Y. [Key Laboratory of Magnetic Levitation Technologies and Maglev Trains (Ministry of Education of China), Superconductivity and New Energy R and D Center, Southwest Jiaotong University, Chengdu 610031 (China); Li, C.S. [Superconducting Materials Center, Northwest Institute for Nonferrous Metal Research, Xi’an 710016 (China); Liu, X.H. [National Engineering Laboratory for Superconducting Material, Western Superconducting Technologies (WST) Co., Ltd., Xi’an 710018 (China); Feng, Y.; Zhang, P.X. [National Engineering Laboratory for Superconducting Material, Western Superconducting Technologies (WST) Co., Ltd., Xi’an 710018 (China); Key Laboratory of Magnetic Levitation Technologies and Maglev Trains (Ministry of Education of China), Superconductivity and New Energy R and D Center, Southwest Jiaotong University, Chengdu 610031 (China); Liu, H.J. [Institute of Plasma Physics, Chinese Academy of Sciences (CAS), Hefei 230031 (China); and others

2014-07-15

Highlights: • This paper reported superconducting properties of the powder-in-tube Nb{sub 3}Al wires. • The Nb{sub 3}Al wires were made by using Nb(Al){sub ss} supersaturated solid solution powders. • The Cu-matrix Nb{sub 3}Al superconducting wires have been successfully fabricated. • The transport J{sub c} of Nb{sub 3}Al wires at 4.2 K, 10 T is up to 12,700 A/cm{sup 2}. - Abstract: High-performance Nb{sub 3}Al superconducting wire is a promising candidate to the application of high-field magnets. However, due to the production problem of km-grade wires that are free from low magnetic field instability, the Nb{sub 3}Al wires made by rapid heating, quenching and transformation (RHQT) are still not available to the large-scale engineering application. In this paper, we reported the properties of the in situ powder-in-tube (PIT) Nb{sub 3}Al superconducting wires, which were made by using the mechanically alloyed Nb(Al){sub ss} supersaturated solid solution, as well as the low temperature heat-treatment at 800 °C for 10 h. The results show that Nb{sub 3}Al superconductors in this method possess very fine grains and well superconducting properties, though a little of Nb{sub 2}Al and Nb impurities still keep being existence at present work. At the Nb{sub 3}Al with a nominal 26 at.% Al content, the onset T{sub c} reaches 15.8 K. Furthermore, a series of Nb{sub 3}Al wires and tapes with various sizes have been fabricated; for the 1.0 mm-diameter wire, the J{sub c} at 4.2 K, 10 T and 14 T have achieved 12,700 and 6900 A/cm{sup 2}, respectively. This work suggests it is possible to develop high-performance Cu-matrix Nb{sub 3}Al superconducting wires by directly using the Nb(Al){sub ss} supersaturated solid-solution without the complex RHQT heat-treatment process.

Effect of tubing condensate on non-invasive positive pressure ventilators tested under simulated clinical conditions.

Science.gov (United States)

Hart, Diana Elizabeth; Forman, Mark; Veale, Andrew G

2011-09-01

Water condensate in the humidifier tubing can affect bi-level ventilation by narrowing tube diameter and increasing airflow resistance. We investigated room temperature and tubing type as ways to reduce condensate and its effect on bi-level triggering and pressure delivery. In this bench study, the aim was to test the hypothesis that a relationship exists between room temperature and tubing condensate. Using a patient simulator, a Res-med bi-level device was set to 18/8 cm H(2)O and run for 6 h at room temperatures of 16°C, 18°C and 20°C. The built-in humidifier was set to a low, medium or high setting while using unheated or insulated tubing or replaced with a humidifier using heated tubing. Humidifier output, condensate, mask pressure and triggering delay of the bi-level were measured at 1 and 6 h using an infrared hygrometer, metric weights, Honeywell pressure transducer and TSI pneumotach. When humidity output exceeded 17.5 mg H(2)O/L, inspiratory pressure fell by 2-15 cm H(2)O and triggering was delayed by 0.2-0.9 s. Heating the tubing avoided any such ventilatory effect whereas warmer room temperatures or insulating the tubing were of marginal benefit. Users of bi-level ventilators need to be aware of this problem and its solution. Bi-level humidifier tubing may need to be heated to ensure correct humidification, pressure delivery and triggering.

Failure Pressure Estimates of Steam Generator Tubes Containing Wear-type Defects

International Nuclear Information System (INIS)

Yoon-Suk Chang; Jong-Min Kim; Nam-Su Huh; Young-Jin Kim; Seong Sik Hwang; Joung-Soo Kim

2006-01-01

It is commonly requested that steam generator tubes with defects exceeding 40% of wall thickness in depth should be plugged to sustain all postulated loads with appropriate margin. The critical defect dimensions have been determined based on the concept of plastic instability. This criterion, however, is known to be too conservative for some locations and types of defects. In this context, the accurate failure estimation for steam generator tubes with a defect draws increasing attention. Although several guidelines have been developed and are used for assessing the integrity of defected tubes, most of these guidelines are related to stress corrosion cracking or wall-thinning phenomena. As some of steam generator tubes are also failed due to fretting and so on, alternative failure estimation schemes for relevant defects are required. In this paper, three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of steam generator tubes with different defect configurations; elliptical wastage type, wear scar type and rectangular wastage type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of the steam generator tube. After investigating the effect of key parameters such as wastage depth, wastage length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wastage region. Comparison of failure pressures predicted according to the proposed estimation scheme with some corresponding burst test data shows good agreement, which provides a confidence in the use of the proposed equations to assess the integrity of steam generator tubes with wear-type defects. (authors)

Hierarchically structured nanoporous carbon tubes for high pressure carbon dioxide adsorption

Directory of Open Access Journals (Sweden)

Julia Patzsch

2017-05-01

Full Text Available Mesoscopic, nanoporous carbon tubes were synthesized by a combination of the Stoeber process and the use of electrospun macrosized polystyrene fibres as structure directing templates. The obtained carbon tubes have a macroporous nature characterized by a thick wall structure and a high specific surface area of approximately 500 m²/g resulting from their micro- and mesopores. The micropore regime of the carbon tubes is composed of turbostratic graphitic areas observed in the microstructure. The employed templating process was also used for the synthesis of silicon carbide tubes. The characterization of all porous materials was performed by nitrogen adsorption at 77 K, Raman spectroscopy, infrared spectroscopy, thermal gravimetric analysis (TGA, scanning electron microscopy (SEM as well as transmission electron microscopy (TEM. The adsorption of carbon dioxide on the carbon tubes at 25 °C at pressures of up to 30 bar was studied using a volumetric method. At 26 bar, an adsorption capacity of 4.9 mmol/g was observed. This is comparable to the adsorption capacity of molecular sieves and vertically aligned carbon nanotubes. The high pressure adsorption process of CO2 was found to irreversibly change the microporous structure of the carbon tubes.

Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

Energy Technology Data Exchange (ETDEWEB)

Barnett, P. C. [Canadian General Electric (Canada)

1968-07-15

The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

Deformation of in-service pressure tubes

International Nuclear Information System (INIS)

Sarce, A.L.

1993-01-01

Candu type nuclear reactor pressure tubes suffer deformations during operation. This are consequences of irradiation growth and creep. By means of a computer code which takes into account the material microstructure, the above mentioned deformations are calculated, and results are compared with corresponding values measured at Embalse nuclear power plant. The calculations make explicit inclusion of intergranular stresses caused by an isotropy in the material. (author). 1 ref

A statistical method for draft tube pressure pulsation analysis

International Nuclear Information System (INIS)

Doerfler, P K; Ruchonnet, N

2012-01-01

Draft tube pressure pulsation (DTPP) in Francis turbines is composed of various components originating from different physical phenomena. These components may be separated because they differ by their spatial relationships and by their propagation mechanism. The first step for such an analysis was to distinguish between so-called synchronous and asynchronous pulsations; only approximately periodic phenomena could be described in this manner. However, less regular pulsations are always present, and these become important when turbines have to operate in the far off-design range, in particular at very low load. The statistical method described here permits to separate the stochastic (random) component from the two traditional 'regular' components. It works in connection with the standard technique of model testing with several pressure signals measured in draft tube cone. The difference between the individual signals and the averaged pressure signal, together with the coherence between the individual pressure signals is used for analysis. An example reveals that a generalized, non-periodic version of the asynchronous pulsation is important at low load.

Fuel-element vibration and bearing pad to pressure tube fretting

International Nuclear Information System (INIS)

Fisher, N.J.; Taylor, C.E.; Pettigrew, M.J.

1990-08-01

Fuel channel operation under boiling condition results in increased flow velocities, which may lead to unacceptable fuel-element vibration and bearing pad to pressure tube fretting. The existing endurance test database does not fully cover the range of future channel operating conditions. In particular, after refuelling, some channels for future designs may operate with two-phase flow conditions outside the range of endurance test conditions. Full-scale endurance testing at realistic steam-water conditions involves substantial energy costs. Therefore, fundamental laboratory investigations were conducted to define and endurance test matrix which adequately envelops the future range of operating conditions while minimizing both the number of tests and the energy requirement of individual tests. The main focus of the laboratory investigations was to establish the relationships between: fuel channel flow conditions and fuel-element vibration; and fuel-element vibration and bearing pad to pressure tube fretting. The vibration response of a single fuel element was measured over a wide range of operating conditions covering realistic fuel channel conditions and simulated endurance testing conditions. For higher void fractions, the vibration amplitudes measured in air/water were much higher than in steam/water, while for low void fractions, the amplitudes were similar. The measured amplitudes in steam/water varied very little over the range of temperature and pressure investigated. The effects of temperature, pressure tube oxide thickness, vibration amplitude and bearing pad manufacturer on pressure tube fretting were investigated. The fretting rate is extremely temperature dependent. For vibration amplitudes about three or four times greater than expected in-reactor conditions, peak fretting rates were observed in the 225 to 286 degrees C temperature range. Fretting rates were seven times less at the higher temperatures of 300 and 315 degrees C, and the lower temperatures

Boiling on a tube bundle: heat transfer, pressure drop and flow patterns

International Nuclear Information System (INIS)

Agostini, F.

2008-07-01

The complexity of the two-phase flow in a tube bundle presents important problems in the design and understanding of the physical phenomena taking place. The working conditions of an evaporator depend largely on the dynamics of the two-phase flow that in turn influence the heat exchange and the pressure drop of the system. A characterization of the flow dynamics, and possibly the identification of the flow pattern in the tube bundle, is thus expected to lead to a better understanding of the phenomena and to reveal on the mechanisms governing the tube bundle. Therefore, the present study aims at providing further insights into two-phase bundle flow through a new visualization system able to provide for the first time a view of the flow in the core of a tube bundle. In addition, the measurement of the light attenuation of a laser beam through the two-phase flow and measurement of the high frequency pressure fluctuations with a piezo-electric pressure transducer are used to characterize the flow. The design and the validation of this new instrumentation also provided a method for the detection of dry-out in tube bundles. This was achieved by a laser attenuation technique, flow visualization, and estimation of the power spectrum of the pressure fluctuation. The current investigation includes results for two different refrigerants, R134a and R236fa, three saturations temperatures T sat = 5, 10 and 15 °C, mass velocities ranging from 4 to 40 kg/sm² in adiabatic and diabatic conditions (several heat fluxes). Measurement of the local heat transfer coefficient and two-phase frictional pressure drop were obtained and utilized to improve the current prediction methods. The heat transfer and pressure drop data were supported by extensive characterization of the two-phase flow, which was to improve the understanding of the two-phase flow occurring in tube bundles. (author)

Evaluation of critical current density and residual resistance ratio limits in powder in tube Nb$_{3}$Sn conductors

CERN Document Server

Segal, Christopher; Sung, Zu Hawn; Lee, Peter J; Sailer, Bernd; Thoener, Manfred; Schlenga, Klaud; Ballarino, Amalia; Bottura, Luca; Bordini, Bernardo; Scheuerlein, Christian; Larnalestier, David C

2016-01-01

High critical current density ( Jc) Nb$_{3}$Sn A15 multifilamentary wires require a large volume fraction of small grain (SG), superconducting A15 phase, as well as Cu stabilizer with high Residual Resistance Ratio (RRR) to provide electromagnetic stabilization and protection. In powder-in-tube (PIT) wires the unreacted Nb7.5 wt%Ta outer layer of the tubular filaments acts as a diffusion barrier and protects the interfilamentary Cu stabilizer from Sn contamination. A high RRR requirement generally imposes a restricted A15 reaction heat treatment to prevent localized full reaction of the filament that could allow Sn to reach the Cu. In this study we investigate recent high quality PIT wires that achieve a Jc (12 T, 4.2 K) up to ∼2500 A mm−2 and find that the minimum diffusion barrier thickness decreases as the filament aspect ratio increases from ∼1 in the inner rings of filaments to 1.3 in the outer filament rings. We found that just 2–3 diffusion barrier breaches can degrade RRR from 300 to 150 or le...

Prediction of pressure tube fretting-wear damage due to fuel vibration

International Nuclear Information System (INIS)

Yetisir, M.; Fisher, N.J.

1997-01-01

Fretting marks between fuel bundle bearing pads and pressure tubes have been observed at the inlet end of some Darlington Nuclear Generating Station (NGS) and Bruce NGS fuel channels. The excitation mechanisms that lead to fretting are not fully understood. In this paper, the possibility of bearing pad-to-pressure tube fretting due to turbulence-induced motion of the fuel element is investigated. Numerical simulations indicate that this mechanism by itself is not likely to cause the level of fretting experienced in Darlington and Bruce NGSs. (orig.)

Development of out-of-core concepts for a supercritical-water, pressure-tube reactor

International Nuclear Information System (INIS)

Diamond, W.T.

2010-01-01

One of the Generation IV programs at Chalk River Laboratories has as its prime focus the development of out-of-core concepts for the SuperCritical Water (SCW) pressure tube reactor under development in Canada. A number of technical issues associated with the interface of out-of-core components and the pressure tubes of a SCW pressure tube reactor are being investigated. This article focuses on several aspects of out-of-core components and layouts, building upon concepts that have been developed during the past few years. The efforts are strongly focused on concepts for a fuel channel that can be fabricated with the tight lattice pitch (typically 230 to 250 mm) that may be required for some applications such as utilization of a thorium fuel cycle. It is not practical to adapt concepts with a tight lattice pitch while using the thicker materials required for the higher temperatures and pressures required for supercritical operation. A change in lattice pitch or configuration is required to accommodate the component size increases. This presentation will cover a number of new concepts developed to produce feeders and end fittings for the harsh conditions of a SCW pressure tube reactor. These components are then developed into conceptual models of a Gen IV pressure tube reactor mounted in both horizontal and vertical orientations. Full 3-D solid models of both concepts will be demonstrated as well as a 1/10th-scale model of one face of a horizontal concept that has been built from components made with a 3-D printer. (author)

Integrated probabilistic assessment for DHC initiation, growth and leak-before-break of PHWR pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Oh, Young-Jin [Power Engineering Research Institute, KEPCO Engineering and Construction, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

2014-08-15

Highlights: • We develop an integrated approach for probabilistic assessment of PHWR pressure tube. • We examine probabilities of DHC initiation, growth, penetration and LBB failure. • The proposed approach is helpful to calculate rupture probabilities in reactor flaws even in the case of very low rupture probability. - Abstract: A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor.

Integrated probabilistic assessment for DHC initiation, growth and leak-before-break of PHWR pressure tubes

International Nuclear Information System (INIS)

Oh, Young-Jin; Chang, Yoon-Suk

2014-01-01

Highlights: • We develop an integrated approach for probabilistic assessment of PHWR pressure tube. • We examine probabilities of DHC initiation, growth, penetration and LBB failure. • The proposed approach is helpful to calculate rupture probabilities in reactor flaws even in the case of very low rupture probability. - Abstract: A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor

Characterizing the active opening of the eustachian tube in a hypobaric/hyperbaric pressure chamber.

Science.gov (United States)

Mikolajczak, Stefanie; Meyer, Moritz Friedo; Hahn, Moritz; Korthäuer, Christine; Jumah, Masen Dirk; Hüttenbrink, Karl-Bernd; Grosheva, Maria; Luers, Jan Christoffer; Beutner, Dirk

2015-01-01

Active and passive opening of the Eustachian tube (ET) enables direct aeration of the middle ear and a pressure balance between middle ear and the ambient pressure. The aim of this study was to characterize standard values for the opening pressure (ETOP), the opening frequency (ETOF), and the opening duration (ETOD) for active tubal openings (Valsalva maneuver, swallowing) in healthy participants. In a hypobaric/hyperbaric pressure chamber, 30 healthy participants (19 women, 11 men; mean age, 25.57 ± 3.33 years) were exposed to a standardized profile of compression and decompression. The pressure values were recorded via continuous impedance measurement during the Valsalva maneuver and swallowing. Based on the data, standard curves were identified and the ETOP, ETOD, and ETOF were determined. Recurring patterns of the pressure curve during active tube opening for the Valsalva maneuver and for active swallowing were characterized. The mean value for the Valsalva maneuver for ETOP was 41.21 ± 17.38 mbar; for the ETOD, it was 2.65 ± 1.87 seconds. In the active pressure compensation by swallowing, the mean value for the ETOP was 29.91 ± 13.07 mbar; and for the ETOD, it was 0.82 ± 0.53 seconds. Standard values for the opening pressure of the tube and the tube opening duration for active tubal openings (Valsalva maneuver, swallowing) were described, and typical curve gradients for healthy subjects could be shown. This is another step toward analyzing the function of the tube in compression and decompression.

Design of a high-pressure single pulse shock tube for chemical kinetic investigations

International Nuclear Information System (INIS)

Tranter, R. S.; Brezinsky, K.; Fulle, D.

2001-01-01

A single pulse shock tube has been designed and constructed in order to achieve extremely high pressures and temperatures to facilitate gas-phase chemical kinetic experiments. Postshock pressures of greater than 1000 atmospheres have been obtained. Temperatures greater than 1400 K have been achieved and, in principle, temperatures greater than 2000 K are easily attainable. These high temperatures and pressures permit the investigation of hydrocarbon species pyrolysis and oxidation reactions. Since these reactions occur on the time scale of 0.5--2 ms the shock tube has been constructed with an adjustable length driven section that permits variation of reaction viewing times. For any given reaction viewing time, samples can be withdrawn through a specially constructed automated sampling apparatus for subsequent species analysis with gas chromatography and mass spectrometry. The details of the design and construction that have permitted the successful generation of very high-pressure shocks in this unique apparatus are described. Additional information is provided concerning the diaphragms used in the high-pressure shock tube

Advanced NDE (ANDE) and its application for pressure tube inspections in OPG reactors

Energy Technology Data Exchange (ETDEWEB)

Jarron, D.; Trelinski, M.; Kretz, S. [Ontario Power Generation, Ajax, Ontario (Canada)]. E-mail: don.jarron@opg.com; mike.trelinski@opg.com; steve.kretz@opg.com

2006-07-01

Periodic and in-service inspections of CANDU fuel channels are essential for the proper assessment of the structural integrity of these vital components. The arrival of new delivery devices for fuel channel inspections (Universal Delivery Machine) has driven new methods for gathering and analyzing NDE data. The Advanced Non-Destructive Examination (ANDE) system has been designed and field implemented as a high speed data acquisition system to meet the requirements of the CSA N285.4 code. It was built from the solid foundation of CIGAR experience and uses cutting edge hardware and software to attain high speed data collection enabling relatively quick inspection of a large number of fuel channels. The capabilities of the ANDE inspection system include: Surface and volumetric inspection of pressure tube by ultrasonics; Flaw characterization by ultrasonics; Pressure tube diameter measurements; Pressure tube thickness measurements; Garter Spring location by Eddy Current; Garter Spring location by ultrasonics; Pressure tube sag measurement. In addition to the above, selected flaws/areas of a pressure tube can be replicated using a two plate ANDE replica tool. At the heart of the inspection system is a set of twelve ultrasonic probes positioned in such a way that the inspected areas are examined from various angles and directions and by various ultrasonic wave modes (shear and longitudinal). High frequency ultrasound used for the examinations allows for reliable detection of small flaws. Separate sensors have been installed on the inspection head for Garter Spring location and sag measurements. (author)

Tracheal tube and laryngeal mask cuff pressure during anaesthesia - mandatory monitoring is in need

DEFF Research Database (Denmark)

Rokamp, K.Z.; Secher, N.H.; Møller, Ann

2010-01-01

ABSTRACT: BACKGROUND: To prevent endothelium and nerve lesions, tracheal tube and laryngeal mask cuff pressure is to be maintained at a low level and yet be high enough to secure air sealing. METHOD: In a prospective quality-control study, 201 patients undergoing surgery during anaesthesia (without...... the use of nitrous oxide) were included for determination of the cuff pressure of the tracheal tubes and laryngeal masks. RESULTS: In the 119 patients provided with a tracheal tube, the median cuff pressure was 30 (range 8 - 100) cm H2O and the pressure exceeded 30 cm H2O (upper recommended level) for 54...... patients. In the 82 patients provided with a laryngeal mask, the cuff pressure was 95 (10 - 121) cm H2O and above 60 cm H2O (upper recommended level) for 56 patients and in 34 of these patients, the pressure exceeded the upper cuff gauge limit (120 cm H2O). There was no association between cuff pressure...

Fuel bundle to pressure tube fretting in Bruce and Darlington

Energy Technology Data Exchange (ETDEWEB)

Norsworthy, A G; Ditschun, A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs.

Fuel bundle to pressure tube fretting in Bruce and Darlington

International Nuclear Information System (INIS)

Norsworthy, A.G.; Ditschun, A.

1995-01-01

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs

Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

International Nuclear Information System (INIS)

Majumdar, S.; Kasza, K.

2009-01-01

The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

Majumdar, S.; Kasza, K. [Argonne National Laboratory, Nuclear Energy Division, Lemont, Illinois (United States)

2009-07-01

The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

The EL-4 reactor. Changing of a pressure tube on a test loop

International Nuclear Information System (INIS)

Foulquier, H.; Clara, P.

1964-01-01

Right from the beginning of the EL-4 project, the research convected with the overall design of the reactor was guided by the various technical specifications resulting from a justifiable concern about the reliability. The external and internal tubes of each layer situated in the reactor block had in particular to be interchangeable. The research alone into the dismantling of the external tube, i.e in fact the pressure tube, justified a certain number of full-scale tests on a model. The tests carried out under relevant conditions on a non-irradiated structure made it possible to define a complete ranger of of positioning and un-positioning sequences at a distance for such a pressure tube. (authors) [fr

Experimental and visual study on flow patterns and pressure drops in U-tubes

International Nuclear Information System (INIS)

Da Silva Lima, J. R.

2011-01-01

In single- and two-phase flow heat exchangers (in particular 'coils'), besides the straight tubes there are also many singularities, in particular the 180° return bends (also called return bends or U-bends). However, contrary to the literature concerning pressure drops and heat transfer in straight tubes, where many experimental data and predicting methods are available, only a limited number of studies concerning U-bends can be found. Neither reliable experimental data nor proven prediction methods are available. Indeed, flow structure, pressure drop and heat transfer in U-bends are an old unresolved design problem in the heat transfer industry. Thus, the present study aims at providing further insight on two-phase pressure drops and flows patterns in U-bends. Based on a new type of U-bend test section, an extensive experimental study was conducted. The experimental campaign covered five test sections with three internal diameters (7.8, 10.8 and 13.4 mm), five bend diameters (24.8, 31.7, 38.1, 54.8 and 66.1 mm), tested for three orientations (horizontal, vertical upflow and vertical downflow), two fluids (R134a and R410A), two saturation temperatures (5 and 10 °C) and mass velocities ranging from 150 to 1000 kg s -1 m -2 . The flow pattern observations identified were stratified-wavy, slug-stratified-wavy, intermittent, annular, dryout and mist flows. The effects of the U-bend on the flow patterns were also observed. A total of 5655 pressure drop data were measured at seven different locations in the test section ( straight tubes and U-bend) providing a total of almost 40,000 data points. The straight tube data were first used to improve the actual two-phase straight tube model of Moreno-Quibén and Thome. This updated model was then used to developed a two-phase U-bend pressure drop model. Based on a comparison between experimental and predicted values, it is concluded that the new two-phase frictional pressure drop model for U-bends successfully

Reactive synthesis of NbAl3 matrix composites

International Nuclear Information System (INIS)

Lu, L.; Kim, Y.S.; Gokhale, A.B.; Abbaschian, R.

1990-01-01

NbAl 3 matrix composites were synthesized in-situ via reactive hot compaction (RHC) of elemental powders. It was found that the simultaneous application of pressure during synthesis was effective in attaining a near-theoretical density matrix at relatively low temperatures and pressures. Using this technique, two types of composites were produced: matrices containing a uniform dispersion of second phase particles (either Nb 3 Al or Nb 2 Al with an Nb core or Nb 2 Al) and matrices reinforced with coated or uncoated ductile Nb filaments. It was found that a limited amount of toughening is obtained using the first approach, while composites containing coated Nb filaments exhibited a significant increase in the ambient temperature fracture toughness. In this paper, various aspects of RHC processing of NbAl 3 matrix composites, the effect of initial stoichiometry and powder size on the microstructure, as well as the mechanical behavior of the composites are discussed

Development of NDT techniques for the inspection of WSGHWR pressure tubes

International Nuclear Information System (INIS)

Gray, B.S.; Highmore, P.J.; Rudlin, J.R.; Cooper, A.G.

1979-01-01

The fuel for the Steam Generating Heavy Water Reactor at Winfrith Heath is contained in vertical Zircaloy pressure tubes and is cooled by boiling light water. This paper describes the development of NDT techniques for the inservice examination of the pressure tubes to provide continuing assurance of the absence of axial crack-like defects. The resultant equipment has to operate in water-filled tubes in the presence of the radiation field due to the irradiated fuel elements in adjacent tubes. Also, a layer of surface oxide on the inside of the tubes has been found to significantly affect the behaviour of a prototype inspection device. To provide adequate sensitivity in these conditions, without the occurrence of unnecessary spurious indications, a combination of techniques has been developed. This involves the use of ultrasonics in both pulse-echo and 'pitch and catch' mode together with a single frequency eddy current technique. Laboratory work using artificial defects is described and also how the development programme was modified to accommodate the results of in-reactor tests using a prototype device. Reference is also made to the development of CCTV equipment to provide a supplementary visual examination. (author)

The terminal solid solubility of hydrogen and deuterium in Zr-2.5Nb alloys

Energy Technology Data Exchange (ETDEWEB)

Ritchie, I G; Pan, Z L; Puls, M P [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

1997-02-01

The presence of hydrides in zirconium based alloys is an important factor in assessing the potential for delayed hydride cracking in pressure tubes and the embrittlement of other in-core components fabricated from these alloys. Consequently, the terminal solid solubility (TSS) of hydrogen in the zirconium alloys used in the Nuclear Industry is an important parameter. However, at the low hydrogen concentrations found in practice, the TSS is difficult to measure accurately and even the measurements of hydrogen concentrations by standard techniques are notoriously difficult to make reproducibly at the nominal levels found in pressure tube materials. The presence of hydrides, their dissolution and nucleation gives rise to a number of internal friction phenomena and changes in Young`s modulus that can be useful from the practical point of view. These phenomena can be used to establish expressions for the TSS as a function of temperature, the hysteresis between dissolution and nucleation and hydrogen supercharging from the gas phase. In particular, such studies show that the hysteresis between the TSS measured during heating and cooling is particularly sensitive to the thermal history of the sample. This paper reviews the phenomena involved and presents some recent results on Zr-2.5Nb pressure tube material. (author). 28 refs, 17 figs, 6 tabs.

Boiling on a tube bundle: heat transfer, pressure drop and flow patterns

International Nuclear Information System (INIS)

Royen Van, E.

2011-11-01

The complexity of two-phase flow boiling on a tube bundle presents many challenges to the understanding of the physical phenomena taking place. It is important to quantify these numerous heat flow mechanisms in order to better describe the performance of tube bundles as a function of the operational conditions. In the present study, the bundle boiling facility at the Laboratory of Heat and Mass Transfer (LTCM) was modified to obtain high-speed videos to characterise the two-phase regimes and some bubble dynamics of the boiling process. It was then used to measure heat transfer on single tubes and in bundle boiling conditions. Pressure drop measurements were also made during adiabatic and diabatic bundle conditions. New enhanced boiling tubes from Wolverine Tube Inc. (Turbo-B5) and the Wieland-Werke AG (Gewa-B5) were investigated using R134a and R236fa as test fluids. The tests were carried out at saturation temperatures T sat of 5 °C and 15 °C, mass flow rates from 4 to 35 kg/m 2 s and heat fluxes from 15 to 70 kW/m 2 , typical of actual operating conditions. The flow pattern investigation was conducted using visual observations from a borescope inserted in the middle of the bundle. Measurements of the light attenuation of a laser beam through the intertube two-phase flow and local pressure fluctuations with piezo-electric pressure transducers were also taken to further help in characterising the complex flow. Pressure drop measurements and data reduction procedures were revised and used to develop new, improved frictional pressure drop prediction methods for adiabatic and diabatic two-phase conditions. The physical phenomena governing the enhanced tube evaporation process and their effects on the performance of tube bundles were investigated and insight gained. A new method based on a theoretical analysis of thin film evaporation was used to propose a new correlating parameter. A large new database of local heat transfer coefficients were obtained and then

Cantilever beam test in Zr-2.5%Nb: Comparative study between pipe material obtained by laminating and wire drawing

International Nuclear Information System (INIS)

Gomez, A G; Szieber, W C; Vizcaino, P; Loureiro, N; Bianchi, D R; Banchik, A D

2012-01-01

The Zr-2.5Nb alloy is used in the manufacture of pressure tubes for the CANDU nuclear power reactors. These tubes are subjected to severe service conditions: one o f them, the heavy water corrosion due to the coolant generates release of hydrogen, part of which ingress in the material raising its initial concentration and exposing them to a phenomena referred as delay hydrogen cracking. The results presented in this paper show the performance of a pressure tube of domestic manufacture under conditions of tension and hydrogen content in order to be compared with the behavior of a standard pressure tube in operation in the nuclear power plant. To do this is, cantilever notched and pre cracked samples were hydrided from both kinds of tubes. Each one of these samples was subjected to the cantilever beam test, which consists in a bending test performed in a furnace at 250 o C. Starting from a stress intensity factor Ki which determines the propagation start of the crack, the growth is followed by the acoustic emission technique up to the arrest of the crack by controlling the bending load. This work presents the comparative data such as critical voltages, behavior of hydrides, and DHC parameters from both trials. Although the number of tests is reduced; results show a good performance of the tubes of domestic manufacture (author)

High temperature deformation behavior of gradually pressurized zircaloy-4 tubes

International Nuclear Information System (INIS)

Suzuki, Motoye

1982-03-01

In order to obtain preliminary perspectives on fuel cladding deformation behavior under changing temperature and pressure conditions in a hypothetical loss-of-coolant accident of PWR, a Zircaloy-4 tube burst test was conducted in both air and 99.97% Ar atomospheres. The tubes were directly heated by AC-current and maintained at various temperatures, and pressurized gradually until rupture occurred. Rupture circumferential strains were generally larger in Ar gas than in air and attained a maximum around 1100 K in both atmospheres. Some tube tested in air produced axially-extended long balloons, which proved not to be explained by such properties or ideas as effect of cooling on strain rate, superplasticity, geometrical plastic instability and stresses generated by surface oxide layer. A cause of the long balloon may be obtained in the anisotropy of the material structure. But even a qualitative analysis based on this property can not be made due to insufficient data of the anisotropy. (author)

Two and dimensional heat analysis inside a high pressure electrical discharge tube

International Nuclear Information System (INIS)

Aghanajafi, C.; Dehghani, A. R.; Fallah Abbasi, M.

2005-01-01

This article represents the heat transfer analysis for a horizontal high pressure mercury steam tube. To get a more realistic numerical simulation, heat radiation at different wavelength width bands, has been used besides convection and conduction heat transfer. The analysis for different gases with different pressure in two and three dimensional cases has been investigated and the results compared with empirical and semi empirical values. The effect of the environmental temperature on the arc tube temperature is also studied

In vitro evaluation of the method effectiveness to limit inflation pressure cuffs of endotracheal tubes

Directory of Open Access Journals (Sweden)

Rafael de Macedo Coelho

2016-04-01

Full Text Available ABSTRACT BACKGROUND AND OBJECTIVE: Cuffs of tracheal tubes protect the lower airway from aspiration of gastric contents and facilitate ventilation, but may cause many complications, especially when the cuff pressure exceeds 30 cm H2O. This occurs in over 30% of conventional insufflations, so it is recommended to limit this pressure. In this study we evaluated the in vitro effectiveness of a method of limiting the cuff pressure to a range between 20 and 30 cm H2O. METHOD: Using an adapter to connect the tested tube to the anesthesia machine, the relief valve was regulated to 30 cm H2O, inflating the cuff by operating the rapid flow of oxygen button. There were 33 trials for each tube of three manufacturers, of five sizes (6.5-8.5, using three times inflation (10, 15 and 20 s, totaling 1485 tests. After inflation, the pressure obtained was measured with a manometer. Pressure >30 cm H2O or <20 cm H2O were considered failures. RESULTS: There were eight failures (0.5%, 95% CI: 0.1-0.9%, with all by pressures <20 cm H2O and after 10 s inflation (1.6%, 95% CI: 0 5-2.7%. One failure occurred with a 6.5 tube (0.3%, 95% CI: -0.3 to 0.9%, six with 7.0 tubes (2%, 95% CI: 0.4-3.6%, and one with a 7.5 tube (0.3%, 95% CI: -0.3 to 0.9%. CONCLUSION: This method was effective for inflating tracheal tube cuffs of different sizes and manufacturers, limiting its pressure to a range between 20 and 30 cm H2O, with a success rate of 99.5% (95% CI: 99.1-99.9%.

Method of reactivity control in pressure tube reactor

International Nuclear Information System (INIS)

Fukumura, Nobuo.

1988-01-01

Purpose: To provide a method of controlling reactivity in a pressure tube reactor at high conversion ratio intended for high burn-up degree. Method: Control tubes are inserted in heavy water moderator. Light water is filled in the tubes at the initial burning stage. Along with the advance of the burning, the light water is gradually removed and replaced with gases of less reactive nuclear reactivity with neutrons such as air or gaseous carbon dioxide. The tubes are made of less neutron absorbing material such as aluminum. By filling light water, infinite multiplication factor is reduced to suppress the reactivity at the initial burning stage. As light water is gradually removed and replaced with air, etc., it provides an effect like that elimination of heavy water moderator to increase the conversion ratio. Accordingly, nuclear fission materials are produced additionally by so much to extend the burn-up degree. In this way, it can provide excellent effect in realizing high burn-up ratio and high conversion ratio. (Kamimura, M.)

Manufacture of thin-walled clad tubes by pressure welding of roll bonded sheets

Science.gov (United States)

Schmidt, Hans Christian; Grydin, Olexandr; Stolbchenko, Mykhailo; Homberg, Werner; Schaper, Mirko

2017-10-01

Clad tubes are commonly manufactured by fusion welding of roll bonded metal sheets or, mechanically, by hydroforming. In this work, a new approach towards the manufacture of thin-walled tubes with an outer diameter to wall thickness ratio of about 12 is investigated, involving the pressure welding of hot roll bonded aluminium-steel strips. By preparing non-welded edges during the roll bonding process, the strips can be zip-folded and (cold) pressure welded together. This process routine could be used to manufacture clad tubes in a continuous process. In order to investigate the process, sample tube sections with a wall thickness of 2.1 mm were manufactured by U-and O-bending from hot roll bonded aluminium-stainless steel strips. The forming and welding were carried out in a temperature range between RT and 400°C. It was found that, with the given geometry, a pressure weld is established at temperatures starting above 100°C. The tensile tests yield a maximum bond strength at 340°C. Micrograph images show a consistent weld of the aluminium layer over the whole tube section.

Effect of Ovality on Maximum External Pressure of Helically Coiled Steam Generator Tubes with a Rectangular Wear

Energy Technology Data Exchange (ETDEWEB)

Shin, Dong In; Lim, Eun Mo; Huh, Nam Su [Seoul National Univ. of Science and Technology, Seoul (Korea, Republic of); Choi, Shin Beom; Yu, Je Yong; Kim, Ji Ho; Choi, Suhn [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

A structural integrity of steam generator tubes of nuclear power plants is one of crucial parameters for safe operation of nuclear power plants. Thus, many studies have been made to provide engineering methods to assess integrity of defective tubes of commercial nuclear power plants considering its operating environments and defect characteristics. As described above, the geometric and operating conditions of steam generator tubes in integral reactor are significantly different from those of commercial reactor. Therefore, the structural integrity assessment of defective tubes of integral reactor taking into account its own operating conditions and geometric characteristics, i. e., external pressure and helically coiled shape, should be made to demonstrate compliance with the current design criteria. Also, ovality is very specific characteristics of the helically coiled tube because it is occurred during the coiling processes. The wear, occurring from FIV (Flow Induced Vibration) and so on, is main degradation of steam generator tube. In the present study, maximum external pressure of helically coiled steam generator tube with wear is predicted based on the detailed 3-dimensional finite element analysis. As for shape of wear defect, the rectangular shape is considered. In particular, the effect of ovality on the maximum external pressure of helically coiled tubes with rectangular shaped wear is investigated. In the present work, the maximum external pressure of helically coiled steam generator tube with rectangular shaped wear is investigated via detailed 3-D FE analyses. In order to cover a practical range of geometries for defective tube, the variables affecting the maximum external pressure were systematically varied. In particular, the effect of tube ovality on the maximum external pressure is evaluated. It is expected that the present results can be used as a technical backgrounds for establishing a practical structural integrity assessment guideline of

Time-dependent leak behavior of flawed Alloy 600 tube specimens at constant pressure

Energy Technology Data Exchange (ETDEWEB)

Bahn, Chi Bum, E-mail: bahn@anl.gov [Argonne National Laboratory, Argonne, IL 60439 (United States); Majumdar, Saurin [Argonne National Laboratory, Argonne, IL 60439 (United States); Harris, Charles [United States Nuclear Regulatory Commission, Rockville, MD 20852 (United States)

2011-10-15

Leak rate testing has been performed using Alloy 600 tube specimens with throughwall flaws. Some specimens have shown time-dependent leak behavior at constant pressure conditions. Fractographic characterization was performed to identify the time-dependent crack growth mechanism. The fracture surface of the specimens showed the typical features of ductile fracture, as well as the distinct crystallographic facets, typical of fatigue crack growth at low {Delta}K level. Structural vibration appears to have been caused by the oscillation of pressure, induced by a high-pressure pump used in a test facility, and by the water jet/tube structure interaction. Analyses of the leak behaviors and crack growth indicated that both the high-pressure pump and the water jet could significantly contribute to fatigue crack growth. To determine whether the fatigue crack growth during the leak testing can occur solely by the water jet effect, leak rate tests at constant pressure without the high-pressure pump need to be performed. - Highlights: > Leak rate of flawed Alloy 600 tubing increased at constant pressure condition. > Fractography revealed two cases: ductile tearing and crystallographic facets. > Crystallographic facets are typical features of fatigue crack growth at low {Delta}K. > Fatigue source could be water jet-induced vibration and/or high-pressure pump pulsation.

An assessment of the waterside corrosion and hydrogen pick-up in the zircaloy-2 pressure tubes of PHWR

International Nuclear Information System (INIS)

Sah, D.N.

1992-01-01

In view of the deleterious effect of hydriding on the operating life of zircaloy-2 pressure tubes in PHWRs there is an urgent need for the assessment of the status of the pressure tubes with respect to corrosion and hydrogen pick-up in the operating PHWRs. A model has been developed for analysing the waterside corrosion and hydrogen pick-up in the zircaloy-2 pressure tubes under reactor operating conditions. This model predicts the axial profiles of oxide layer thickness and hydrogen pick-up in the pressure tubes as a function of the operating time of the reactor. The prediction of hydrogen pick-up by the model in the F-10 pressure tube of RAPS-I have been found to be in good agreement with the measured value of hydrogen content. This report gives a brief description of the model and its predictions on the present status of hydrogen pick-up in the pressure tubes of lead reactor RAPS-II. (author). 6 refs., 5 figs., 2 tabs

Prediction of pressure tube fretting-wear damage due to fuel vibration

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M; Fisher, N J [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

1996-12-31

Fretting marks between fuel bundle bearing pads and pressure tubes have been observed at the inlet end of some Darlington NGS (nuclear generating station) and Bruce NGS fuel channels. The excitation mechanisms that lead to fretting are not fully understood. In this paper, the possibility of bearing pad-to-pressure tube fretting due to turbulence-induced motion of the fuel element is investigated. Numerical simulations indicate that this mechanism by itself is not likely to cause the level of fretting experienced in Darlington and Bruce NGS`s (nuclear generating stations). (author). 12 refs., 2 tabs., 11 figs.

Study on frictional pressure drop of steam-water two phase flow in optimized four-head internal-ribbed tube

International Nuclear Information System (INIS)

Wang Weishu; Zhu Xiaojing; Bi Qincheng; Wu Gang; Yu Shuiqing

2012-01-01

The optimized internal-ribbed tube is different from the normal internal-ribbed tube on the frictional pressure drop characteristics. The frictional pressure drop characteristics of steam-water two phase flow in horizontal four-head optimized internal-ribbed were studied under adiabatic condition. According to the experimental and calculation results, the two-phase multiplier is greatly affected by the steam quality and pressure. The two-phase multiplier increases with increasing quality, and decreases with increasing pressure. In the near-critical pressure region, the two-phase multiplier is close to 1. The frictional pressure drop of two phase flow in optimized tube is less than that in the normal tube under the same work condition. The good hydrodynamic condition could be achieved when the optimized internal-ribbed tube is used in the heat transfer equipment because the self-compensating characteristics exist due to the reduction of frictional pressure drop. (authors)

Hydrogen absorption in epitaxial W/Nb(001) and polycrystalline Fe/Nb(110) multilayers studied in-situ by X-ray/neutron scattering techniques and X-ray absorption spectroscopy

International Nuclear Information System (INIS)

Klose, F.; Rehm, C.; Fieber-Erdmann, M.; Holub-Krappe, E.; Bleif, H. J.; Sowers, H.; Goyette, R.; Troger, L.; Maletta, H.

1999-01-01

Hydrogen can be absorbed in large quantities by 100 A thin Nb layers embedded in epitaxial W/Nb and polycrystalline Fe/Nb multilayers. The solubility and the hydrogen-induced structural changes of the host lattice are explored in-situ by small-angle neutron/X-ray reflectometry and high-angle diffraction. These measurements reveal for both systems that the relative out-of-plane expansion of the Nb layers is considerably larger than the relative increase of the Nb interplanar spacing indicating two distinctly different mechanisms of hydrogen absorption. In Fe/Nb multilayers, hydrogen expands the Nb interplanar spacing in a continuous way as function of the external pressure. In contrast, the Nb lattice expansion is discontinuous in epitaxial W/Nb multilayers: A jump in the Nb(002) Bragg reflection position occurs at a critical hydrogen pressure of 1 mbar. In-situ EXAFS spectroscopy also exhibits an irreversible expansion of the Nb lattice in the film plane for p H2 > 1 mbar. This can be regarded as a structural phase transition from an exclusively out-of-plane to a three-dimensionally expanded state at low and high hydrogen pressures, respectively

Detection of steam generator tube leaks in pressurized water reactors

International Nuclear Information System (INIS)

Roach, W.H.

1985-01-01

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

High pressure thimble/guide tube seal fitting with built-in low pressure seal especially suitable for facilitated and more efficient nuclear reactor refueling service

International Nuclear Information System (INIS)

Bhatt, P.N.; Blaushield, R.M.

1991-01-01

This patent describes a HP/LP seal arrangement for an elongated guide tube and an elongated thimble disposed therein. The guide tube and thimble extending outwardly from the core of a nuclear reactor to a seal table where the guide tube is welded to the seal table to provide a high pressure seal relative thereto. It comprises: a tubular seal fitting disposed in alignment with the guide tube with the thimble extending therethrough on the low pressure side of the seal table; first high pressure sealing means coupling one end of the fitting to an end of the guide tube to prevent leakage from within the guide tube; inwardly facing thread means disposed adjacent the other and outer end of the seal fitting; a nut having an opening through which the thimble extends and further having outwardly facing threading in mating engagement with the fitting thread means; the fitting having a seal seat spaced longitudinally inwardly from the thread means and facing the fitting outer end and further disposed annularly about the inner surface of the fitting; deformable ring seal means; second releasable high pressure sealing means coupling the thimble to the outer end portion of the guide tube

Correlations of CO2 at supercritical pressures in a vertical circular tube

International Nuclear Information System (INIS)

Li Zhihui; Jiang Peixue

2010-01-01

The experiment results of convection heat transfer of CO 2 at supercritical pressures in a 2 mm diameter vertical circular tube for upward flow and downward flow were analyzed for pressures ranging from 78 to 95 bar, inlet temperatures from to 25 to 40 degree C, and inlet Re numbers from 3000 to 20000. The results were compared with some well known empirical correlations for the heat transfer without buoyancy effects and the heat transfer with strong buoyancy effects. It is found that there is a big deviation between the experiment results and empirical correlations. Based on the experiment data, correlations are developed for the local Nusselt correlations of CO 2 at supercritical pressures in vertical circular tubes.(authors)

THE EFFECTS OF AREA CONTRACTION ON SHOCK WAVE STRENGTH AND PEAK PRESSURE IN SHOCK TUBE

Directory of Open Access Journals (Sweden)

A. M. Mohsen

2012-06-01

Full Text Available This paper presents an experimental investigation into the effects of area contraction on shock wave strength and peak pressure in a shock tube. The shock tube is an important component of the short duration, high speed fluid flow test facility, available at the Universiti Tenaga Nasional (UNITEN, Malaysia. The area contraction was facilitated by positioning a bush adjacent to the primary diaphragm section, which separates the driver and driven sections. Experimental measurements were performed with and without the presence of the bush, at various diaphragm pressure ratios, which is the ratio of air pressure between the driver (high pressure and driven (low pressure sections. The instantaneous static pressure variations were measured at two locations close to the driven tube end wall, using high sensitivity pressure sensors, which allow the shock wave strength, shock wave speed and peak pressure to be analysed. The results reveal that the area contraction significantly reduces the shock wave strength, shock wave speed and peak pressure. At a diaphragm pressure ratio of 10, the shock wave strength decreases by 18%, the peak pressure decreases by 30% and the shock wave speed decreases by 8%.

The Sensitivity Analysis of Axial Pressure Tube Creep Profile for Dryout Power in PHWR

Energy Technology Data Exchange (ETDEWEB)

Ryu, Euiseung; Kim, Youngae [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Stern Laboratory performed the CHF tests with only one axial pressure tube creep profile per 3.3%, 5.1% peak crept channel and made CHF correlation including creep factor from the CHF test results. Wolsong nuclear power plants also have utilized the same CHF correlation derived by CNL. Pressure tube diameter creep rate is function of fast neutron, coolant temperature, and coolant pressure in a channel. It means that various axial pressure tube creep profiles exist in PHWR due to the history of operating conditions. Usually, CHF correlation is used during ROP(Regional Overpower Protection) Trip Setpoint Analysis or Safety Analysis in PHWR. The sensitivity analysis for CHF effects using various creep profiles is needed. This paper summarizes the comparison results of dryout power between CHF test creep profile and estimated creep profiles of Wolsong units. The effect of axial pressure tube creep profile for dryout power in fuel channel is evaluated by using Stern Lab. CHF test creep profile and 380 channel creep profiles of Wolsong. The dryout powers at 3.3% and 5.1% test conditions are slightly smaller when using 380 Wolsong channels creep profiles. These also show that the simulated dryout powers maintain consistency regardless of flow conditions.

Flow instability research on steam generator with straight double-walled heat transfer tube for FBR. Pressure drop under high pressure condition

International Nuclear Information System (INIS)

Liu, Wei; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

2008-01-01

For the Steam Generator (SG) with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important items need researching. As the first step of the research, thermal hydraulics experiments were performed under high pressure condition in JAEA with using a straight tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper evaluates the pressure drop data with TRAC-BF1 code. The Pffan's correlation for single phase flow and the Martinelli-Nelson's two-phase flow multiplier are found can be well predicted the present pressure drop data under high pressure condition. (author)

Burst pressure and leak rate from fretted SG tubes

International Nuclear Information System (INIS)

Hwang, Seong Sik; Jung, Man Kyo; Kim, Hong Pyo; Kim, Joung Soo

2005-01-01

Steam generator(SG) tubes of a pressurized water reactor(PWR) have suffered from various types of corrosion, such as pitting, wastage and stress corrosion cracking (SCC) on both the primary and secondary side. Recently, fretting/wear degradation at the tube support region has been reported in some Korean nuclear power plants. In order to prevent the primary coolant from leaking to the secondary side, the tubes are repaired by a sleeving or plugging. It is important to establish the repair criteria to assure a reactor integrity and yet maintain the plugging ratio within the limits needed for an efficient operation. The objective of the burst test is to obtain a relationship between the burst/leak rate and the shape of the fretted flaws machined with an electro discharge machining (EDM)

Brittle-fracture potential of irradiated Zircaloy-2 pressure tubes

Science.gov (United States)

Huang, F. H.

1993-12-01

Neutron irradiation can degrade the fracture toughness of Zircaloy-2 and may cause highly irradiated reactor components of this material to fail in a brittle manner. The effects of radiation embrittlement on the structural integrity of N Reactor pressure tubes are studied by performing KIc and JIc fracture toughness testing on samples cut from the Zircaloy-2 tubes periodically removed from the reactor. A fluence of 6 × 10 25n/ m2 ( E > 1 MeV) reduced the fracture toughness of the material by 40 to 50%. The fracture toughness values appear to saturate at 260°C with fluences above 3 × 10 25n/ m2 ( E > 1 MeV), but continue to decline with increasing fluence at temperatures below 177°C. Present and previous results obtained from irradiated pressure tubes indicate that the brittle-fracture potential of Zircaloy-2 increases with decreasing temperature and increasing fluence. Fractographic examinations of the fracture surfaces of irradiated samples reveal that circumferential hydride formation significantly influenced fracture morphology by providing sites for easy crack nucleation and leaving deep cracks. However, the deep cracks created at the hydride platelets in specimens containing less than 220 ppm hydrogen are not believed to be the major cause of degradation in postirradiation fracture toughness.

Plane strain analytical solutions for a functionally graded elastic-plastic pressurized tube

International Nuclear Information System (INIS)

Eraslan, Ahmet N.; Akis, Tolga

2006-01-01

Plane strain analytical solutions to functionally graded elastic and elastic-plastic pressurized tube problems are obtained in the framework of small deformation theory. The modulus of elasticity and the uniaxial yield limit of the tube material are assumed to vary radially according to two parametric parabolic forms. The analytical plastic model is based on Tresca's yield criterion, its associated flow rule and ideally plastic material behaviour. Elastic, partially plastic and fully plastic stress states are investigated. It is shown that the elastoplastic response of the functionally graded pressurized tube is affected significantly by the material nonhomogeneity. Different modes of plasticization may take place unlike the homogeneous case. It is also shown mathematically that the nonhomogeneous elastoplastic solution presented here reduces to that of a homogeneous one by appropriate choice of the material parameters

Heat transfer test in a tube using CO2 at supercritical pressures

International Nuclear Information System (INIS)

Kim, Hwan Yeol; Kim, Hyungrae; Song, Jin Ho; Cho, Bong Hyun; Bae, Yoon Yeong

2005-01-01

Heat transfer test facility, which is named as SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt Generation), has been constructed in KAERI for the study of heat transfer and pressure drop characteristics in a single tube, single rod and rod bundle at supercritical CO 2 conditions. The tests with supercritical water are difficult it terms of cost and effort, since the critical pressure and temperature of water are as high as 22.12 MPa and 374.14degC. As a substitute for water, CO 2 is selected for the test since the critical pressure and temperature of CO 2 are 7.38 MPa and 31.05degC that are much lower than those of water. This paper describes the design characteristics of the SPHINX and the experimental investigations on the heat transfer and pressure drop of a vertical single tube with an inside diameter of 4.4 mm with upward flow of supercritical CO 2 . The geometry of the single tube is the same as that of Kyushu University test performed with Freon (R22) for the direct comparison of a medium effect. The tests were performed with various heat and mass fluxes at a given pressure. The range of mass flux is 400∼1200 kg/m 2 s and the heat flux is chosen up to 150 kW/m 2 . The selected pressure are 7.75, 8.12, and 8.85 MPa. The test results are investigated and compared with the previous tests. (author)

Investigation of structural changes in chiral magnet Cr{sub 1∕3}NbS{sub 2} under application of pressure

Energy Technology Data Exchange (ETDEWEB)

Mito, M., E-mail: mitoh@mns.kyutech.ac.jp; Tsuruta, K.; Deguchi, H. [Graduate School of Engineering, Kyushu Institute of Technology, Kitakyushu 804-8550 (Japan); Tajiri, T. [Faculty of Science, Fukuoka University, Fukuoka 814-0180 (Japan); Kishine, J. [Graduate School of Arts and Sciences, The Open University of Japan, Chiba 261-8586 (Japan); Inoue, K. [Graduate School of Science, Hiroshima University, Higashihiroshima 739-8526 (Japan); Institute for Advanced Materials Research, Hiroshima University, Higashihiroshima 739-8526 (Japan); Center for Chiral Science, Hiroshima University, Higashihiroshima 739-8526 (Japan); Kousaka, Y. [Graduate School of Science, Hiroshima University, Higashihiroshima 739-8526 (Japan); Center for Chiral Science, Hiroshima University, Higashihiroshima 739-8526 (Japan); Nakao, Y. [Department of Physics and Mathematics, Aoyama Gakuin University, Sagamihara 229-8558 (Japan); Akimitsu, J. [Graduate School of Science, Hiroshima University, Higashihiroshima 739-8526 (Japan); Center for Chiral Science, Hiroshima University, Higashihiroshima 739-8526 (Japan); Department of Physics and Mathematics, Aoyama Gakuin University, Sagamihara 229-8558 (Japan)

2015-05-14

We perform structural analysis experiments on the chiral magnet Cr{sub 1∕3}NbS{sub 2}, in which Cr{sup 3+} ions are inserted between hexagonal NbS{sub 2} layers. The noncentrosymmetrical nature of the inserted Cr{sup 3+} appears as a distorted CrS{sub 6} octahedron. Under the application of hydrostatic pressure, the lattice shrinks significantly along the c-axis rather than the a-axis. However, at a pressure P of approximately 3–4 GPa, a kink in the rate of decrease in the lattice parameters is observed, and the slight movement of a Nb atom along the c-axis brings about a decrease in the distortion of the CrS{sub 6} octahedron. This structural change qualitatively suggests a decrease in the strength of the Dzyaloshinskii-Moriya (D-M) interaction. Under hydrostatic pressure, the magnetic ordering temperature T{sub C} decreases, and dT{sub C}/dP exhibits a slight change at around 3 GPa. A series of experiments indicates that the change in the structural symmetry of the CrS{sub 6} octahedron influences the exchange network between Cr{sup 3+} ions as well as the D-M interaction.

Apparatus for inspecting and repairing a pressurized-water reactor's steam generator heat exchanger tubes

International Nuclear Information System (INIS)

Mueller, O.; Roettger, H.; Kasti, H.; Hagen, H.G.

1976-01-01

Described is an apparatus provided for use with a pressurized-water reactor' steam generator having a manifold chamber enclosing the bottom side of a horizontal tube sheet having holes therethrough in which are mounted the tubes of a heat exchanger tube bundle. The manifold chamber has a manhole giving access to the tube's bottom side to permit internal inspection or repair of the tubes by registration of an end of a flexible guide conduit with the tube sheet holes and through which a flexible carrier can be guided for insertion via these holes in the tube sheet and through the tubes extending from the tube sheet's other side

Fuel rod bundles proposed for advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Prodea, Iosif; Catana, Alexandru

2010-01-01

The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

High-temperature strength of Nb-1%Zr alloy for irradiation-capsules inner-shell

International Nuclear Information System (INIS)

Nomura, Yasushi; Nakata, Hirokatsu; Tanaka, Mitsuo; Fukaya, Kiyoshi.

1978-04-01

Coated fuel particles in capsules will be irradiated at about 1600 0 C in JMTR. Nb-1%Zr alloy was chosen for inner shell material of the capsules because of its sufficient strength at 1000 0 C and low induced radioactivity. Nb-1%Zr ingot produced by electron beam melting was formed into seamless tubes by hollowing and swaging, followed by annealing. Creep test in helium flow and tensile test in vacuum were made to examine mechanical strength of the Nb-1%Zr tubes at 1000 0 C. Following are the results; 1) 0.2% yield stress at 1000 0 C is about 6 kg/mm 2 . 2) 3000 hr creep rupture stress at 1000 0 C is about 6 kg/mm 2 . (auth.)

Benchmark of WIMS-IST against MCNP for CANDU pressure tube fast fluxes

International Nuclear Information System (INIS)

Donders, R.E.; Douglas, S.R.

2002-01-01

Pressure tube fast-flux data in CANDU are currently calculated using the multi-group neutron transport code WIMS-IST. In this study, the WIMS-IST fast flux calculations are benchmarked against MCNP calculations (a Monte Carlo particle transport code), over the range of fuel burnup and coolant density in CANDU. The comparison shows good agreement between WIMS and MCNP, with WIMS fast fluxes being 1.5% to 4% lower than the MCNP values. The difference is smallest for fresh fuel, and increases with burnup. The fast flux gradient across the pressure tube (factor of 1.23 from inner edge to outer edge) is accurately calculated by WIMS. When reporting fast fluxes in pressure tubes, these are generally given as >1.000 MeV fluxes. For WIMS, this requires an extra conversion step, since the WIMS ENDF/B libraries do not have a group boundary at 1 MeV. The conversion step is based on a fictitious isotope ONEMEV in the WIMS nuclear data library. The conversion factor in WIMS was found to be about one percent too high. When providing >1 MeV fluxes from WIMS, this partially compensates for the slight under prediction of the fast flux. Pressure tube >1 MeV fluxes from WIMS are therefore 0.5% to 3% lower than MCNP values. To obtain accurate fast flux data, neutron transport calculations must be performed on a critical cell. For this study, all calculations were performed with radial albedo boundary conditions giving a critical cell. This required the use of an albedo version of MCNP, developed at AECL. (author)

The development of an auto-sealing system using an electrically shrinkable tube under a low-pressure condition

Energy Technology Data Exchange (ETDEWEB)

Okano, Yoshihiro; Kitagawa, Takao [NKK Corp, Tsu, Mie (Japan); Shoji, Norio [NKK Corp., Yokohama (Japan); Namioka, Toshiyuki [Nippon Kokan Koji Corp., Yokohama (Japan). Research and Development Dept.; Komura, Minoru [Nitto Denko Corp., Fukaya, Saitama (Japan)

1997-04-01

This article describes the development of a system to create high quality, automatic sealing of field joints of polyethylene coated pipelines. The system uses a combination of an electrically heated shrinkable tube and a low-pressure chamber. The self-heating shrinkable tube includes electric heater wires that heat when connected to electricity. A method was developed to eliminate air trapped between the tube and the steel pipe by shrinking the tube under a low-pressure condition. The low-pressure condition was automatic and easily attained by using a vacuum chamber. It was verified that the system produced high quality sealing of the field joints.

Corrosion and hydriding behaviour of some Zr 2.5 wt% Nb alloys in water, steam and various gases at high temperature

Energy Technology Data Exchange (ETDEWEB)

Dalgaard, S. B.

1962-05-15

Fuel sheaths and pressure tubes in Canadian power reactors are at present made from Zircaloy-2. Mechanical properties of a suitably heat treated Zr 2.5 wt% Nb alloy are superior to those of Zircaloy-2, but any new alloy must have resistance to corrosion and hydriding by the coolant and by the gas that insulates the pressure tube from the cold moderator. Exposed to water at temperatures up to 325{sup o}C, the Zr 2.5 wt% Nb alloy has corrosion resistance acceptable for power reactors. Resistance to air and carbon dioxide is less favourable. Addition of tin, or iron and chromium, to the base alloy have little effect on the corrosion resistance, but the addition of copper reduces corrosion in water and steam to some extent and in air and carbon dioxide to a greater extent. Studies of the effect of heat treatment suggest that the amount of niobium in a solid-solution controls the rate of oxidation and hydriding and that concentration, size and distribution of second phase is of little importance. Initial results obtained in NRX indicate that a thermal flux of 3-7 x 10{sup 13} n/cm{sup 2}/sec has little or no effect on oxidation and hydriding in high temperature water. (author)

Numerical study on turbulent heat transfer and pressure drop of nanofluid in coiled tube-in-tube heat exchangers

International Nuclear Information System (INIS)

Aly, Wael I.A.

2014-01-01

Highlights: • The performance of helically coiled tube heat exchanger using nanofluid is modeled. • The 3D turbulent flow and conjugate heat transfer of CTITHE are solved using FVM. • The effects of nanoparticle concentration and curvature ratio are investigated. • The Gnielinski correlation for Nu for turbulent flow in helical tubes can be used for water-based Al 2 O 3 nanofluid. - Abstract: A computational fluid dynamics (CFD) study has been carried out to study the heat transfer and pressure drop characteristics of water-based Al 2 O 3 nanofluid flowing inside coiled tube-in-tube heat exchangers. The 3D realizable k–ε turbulent model with enhanced wall treatment was used. Temperature dependent thermophysical properties of nanofluid and water were used and heat exchangers were analyzed considering conjugate heat transfer from hot fluid in the inner-coiled tube to cold fluid in the annulus region. The overall performance of the tested heat exchangers was assessed based on the thermo-hydrodynamic performance index. Design parameters were in the range of; nanoparticles volume concentrations 0.5%, 1.0% and 2.0%, coil diameters 0.18, 0.24 and 0.30 m, inner tube and annulus sides flow rates from 2 to 5 LPM and 10 to 25 LPM, respectively. Nanofluid flows inside inner tube side or annular side. The results obtained showed a different behavior depending on the parameter selected for the comparison with the base fluid. Moreover, when compared at the same Re or Dn, the heat transfer coefficient increases by increasing the coil diameter and nanoparticles volume concentration. Also, the friction factor increases with the increase in curvature ratio and pressure drop penalty is negligible with increasing the nanoparticles volume concentration. Conventional correlations for predicting average heat transfer and friction factor in turbulent flow regime such as Gnielinski correlation and Mishra and Gupta correlation, respectively, for helical tubes are also valid for

Study on transport safety of fresh MOX fuel. Performance of the cladding tube of fresh MOX fuel against external water pressure

International Nuclear Information System (INIS)

Ito, Chihiro

1999-01-01

It is important to know the ability of the cladding tube for fresh MOX fuel against external water pressure when they were hypothetically sunk into the sea for unknown reasons. In order to evaluate the ability of cladding tubes for MOX fresh fuel against external water pressure, external water pressure tests were carried out. Resistible limit of cladding tubes against external water pressure is defined when cladding tubes are deformed largely due to buckling etc. The test results show cladding tube of BWR type can resist an external water pressure of 69 MPa (a depth of water of 7,000 m) and that of PWR type fuel can resist an external water pressure of 54 MPa (a depth of water of 5,500 m). Moreover, leak tightness is maintained at an external water pressure of 73 MPa (a depth of water of 7,400 m) for BWR type cladding tubes and at an external water pressure of 98 MPa (a depth of water of 10,000 m) for PWR type cladding tubes. (author)

Endotracheal tube cuff pressures during general anaesthesia while using air versus a 50% mixture of nitrous oxide and oxygen as inflating agents

Directory of Open Access Journals (Sweden)

Jesni Joseph Manissery

2007-01-01

Full Text Available The present study was aimed at assessing the efficacy of filling a 50% mixture of nitrous oxide : oxygen (50%N 2 O:O 2 in the endotracheal tube cuff to provide stable cuff pressures during general anaesthesia with 67%N 2 O. The endotracheal tube cuff pressures with air (control as the inflating agent in the tubes were found to have a total mean pressure of 62.60±12.33 at the end of one hour of general anaesthesia. When comparing the endotracheal tube cuff pressures in the Mallinckrodt tubes with that of the Portex tubes, with air as the inflating agent, the Portex tubes showed a significantly lower cuff pressures at the end of one hour. The endotracheal tube cuff pressures with 50%N 2 O:O 2 as the inflating agent showed a total mean pressure of 27.63 ± 3.221 at the end of one hour of general anaesthesia. This indicates that inflation of the cuff of the endotracheal tubes with a 50%N 2 O:O 2 rather than air maintains a stable intra cuff pressure. Therefore, the method of using a 50%N 2 O:O 2 for filling endotracheal tube cuff can be adopted for endotracheal tubes with high-volume, low-pressure cuffs to prevent both excessive cuff pressure and disruption of cuff seal, during general anaesthesia lasting up to one hour.

Delayed Hydride Cracking Mechanism in Zirconium Alloys and Technical Requirements for In-Service Evaluation of Zr-2.5Nb Tubes with Flaws

International Nuclear Information System (INIS)

Kim, Young Suk

2007-01-01

In association with periodic inspection of CANDU nuclear power plant components, Canadian Standards Association issued CSA N285.8 in 2005 as technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors. This first version, CSA N285.8 involves procedures for, firstly, the evaluation of pressure tube flaws, secondly, the evaluation of pressure tube to calandria tube contact and, thirdly, the assessment of a reactor core, and material properties and derived quantities. The evaluation of pressure tube flaws includes delayed hydride cracking evaluation the procedures of which are stipulated based on the existing delayed hydride cracking models. For example, the evaluation of flaw-tip hydride precipitation during reactor cooldown involves a procedure to calculate the equilibrium hydrogen equivalent concentration in solution at the flaw tip, Htipas follows: Htip=Hfexp[- (VH delta no.)/RT], where Hf is the total bulk hydrogen equivalent concentration, VH partial molar volume of hydrogen in zirconium, δ a difference in hydrostatic stress between the bulk and the crack tip. When Htip ≥TSSP at temperature, then flaw-tip hydride is predicted to precipitate. Eq. (1) suggests that hydrogen concentration at the crack tip would increase due to an work energy given by the difference in the hydrostatic stress

Gas pressure in bubble attached to tube circular outlet

OpenAIRE

Salonen, A; Gay, Cyprien; Maestro, A; Drenckhan, W; Rio, Emmanuelle

2016-01-01

In the present Supplementary notes to our work ``Arresting bubble coarsening: A two-bubble experiment to investigate grain growth in presence of surface elasticity'' (accepted in EPL), we derive the expression of the gas pressure inside a bubble located above and attached to the circular outlet of a vertical tube.

Nb46, 5wt% Ti Eb-melting for AC and DC superconducting applications

International Nuclear Information System (INIS)

Bormio, C.; Ramos, M.J.; Pinatti, D.G.

1990-01-01

This paper reports on the superconductor alloy Nb46, 5wt % Ti which presents the best superconducting and mechanical properties for the systems Nb-Ti. The greatest difficulty in obtaining this alloy is related to the difference between the raw materials melting temperatures, which is about 700 degrees C. As a result the alloy homogeneity as well as Ti desired content, turn to be hard to control. The authors choose an electrode sandwich type, where Nb and Ti sheets are interposed. The electrode dimensions calculation is based on the Ti evaporation rate, energy balance and superficial tension of liquid titanium between Nb sheets. The ingots were electron beam melted. Herein, we present the following ingot results: Ti, intersticial and trace contents compared to international manufactures as well as its mechanical workability. This alloy will be used in NbTi wire production for AC and DC applications. The AC and DC wires are produced by coswaging and codrawing of NbTi bars and C u Ni-tubes for AC wires and Cu-tubes for DC wires. High area reductions of about 2 x 10 8 are reached without intermediate heat treatment, and they are essential since they are precursors of collective pinning centers, responsible for high critical current densities

The application of ductile-fracture analysis to predictions of pressure-tube failure

International Nuclear Information System (INIS)

Simpson, L.A.

1981-08-01

Progress during the past six years towards establishing a method for predicting critical crack length in a reactor pressure tube, based on data from tests on small fracture-mechanics specimens, is reviewed. The disadvantages of relying on data from burst tests alone are described along with the benefits of a small-specimen method. It is clear from the work reviewed that only an approach that can account for the ability of the presssure tube material to increase its crack-growth resistance during stable crack extension is suitable for the prediction of critical crack length. A method that utilizes crack-growth resistance curves based on crack-opening displacement, or the J integral, is described, along with a large body of experimental data. It is concluded that the resistance curve approach provides a viable method for the analysis of fracture in pressure tubes that can greatly improve our understanding of the material's behaviour

Bulge testing of copper and niobium tubes for hydroformed RF cavities

Energy Technology Data Exchange (ETDEWEB)

Kim, H.S., E-mail: kim.3237@osu.edu [Department of Materials Science and Engineering, The Ohio State University, Columbus, OH (United States); Sumption, M.D. [Department of Materials Science and Engineering, The Ohio State University, Columbus, OH (United States); Susner, M.A. [Department of Materials Science and Engineering, The Ohio State University, Columbus, OH (United States); Oak Ridge National Laboratory, Oak Ridge, TN (United States); Lim, H. [Department of Materials Science and Engineering, The Ohio State University, Columbus, OH (United States); Sandia National Laboratories, Albuquerque, NM (United States); Collings, E.W. [Department of Materials Science and Engineering, The Ohio State University, Columbus, OH (United States)

2016-01-27

The heat treatment, tensile testing, and bulge testing of Cu and Nb tubes has been carried out to gain experience for the subsequent hydroforming of Nb tube into seamless superconducting radio frequency (SRF) cavities for high energy particle acceleration. In the experimental part of the study samples removed from representative tubes were prepared for heat treatment, tensile testing, residual resistance ratio measurement, and orientation imaging electron microscopy (OIM). After being optimally heat treated Cu and Nb tubes were subjected to hydraulic bulge testing and the results analyzed. In the final part of the study finite-element models (FEM) incorporating constitutive (stress–strain) relationships analytically derived from the tensile and bulge tests, respectively, were used to replicate the bulge test. As expected, agreement was obtained between the experimental bulge parameters and the FEM model based on the bulge-derived constitutive relationship. Not so for the FEM model based on tensile-test data. It is concluded that a constitutive relationship based on bulge testing is necessary to predict a material's performance under hydraulic deformation.

Bulge testing of copper and niobium tubes for hydroformed RF cavities

International Nuclear Information System (INIS)

Kim, H.S.; Sumption, M.D.; Susner, M.A.; Lim, H.; Collings, E.W.

2016-01-01

The heat treatment, tensile testing, and bulge testing of Cu and Nb tubes has been carried out to gain experience for the subsequent hydroforming of Nb tube into seamless superconducting radio frequency (SRF) cavities for high energy particle acceleration. In the experimental part of the study samples removed from representative tubes were prepared for heat treatment, tensile testing, residual resistance ratio measurement, and orientation imaging electron microscopy (OIM). After being optimally heat treated Cu and Nb tubes were subjected to hydraulic bulge testing and the results analyzed. In the final part of the study finite-element models (FEM) incorporating constitutive (stress–strain) relationships analytically derived from the tensile and bulge tests, respectively, were used to replicate the bulge test. As expected, agreement was obtained between the experimental bulge parameters and the FEM model based on the bulge-derived constitutive relationship. Not so for the FEM model based on tensile-test data. It is concluded that a constitutive relationship based on bulge testing is necessary to predict a material's performance under hydraulic deformation.

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2011-07-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

International Nuclear Information System (INIS)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

2011-01-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

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.

Comparative Raman study of Weyl semimetals TaAs, NbAs, TaP and NbP

International Nuclear Information System (INIS)

Liu, H W; Richard, P; Zhao, L X; Chen, G-F; Ding, H

2016-01-01

We report a comparative polarized Raman study of Weyl semimetals TaAs, NbAs, TaP and NbP. The evolution of the phonon frequencies with the sample composition allows us to determine experimentally which atoms are mainly involved for each vibration mode. Our results confirm previous first-principles calculations indicating that the A 1 , B 1 (2), E(2) and E(3) modes involve mainly the As(P) atoms, the B 1 (1) mode is mainly related to Ta(Nb) atoms, and the E(1) mode involves both kinds of atoms. By comparing the energy of the different modes, we establish that the B 1 (1), B 1 (2), E(2) and E(3) become harder with increasing chemical pressure. This behaviour differs from our observation on the A 1 mode, which decreases in energy, in contrast to its behaviour under external pressure. (paper)

Experimental study on condensation heat transfer enhancement and pressure drop penalty factors in four microfin tubes

Energy Technology Data Exchange (ETDEWEB)

Han, D [Korea University, Seoul (Korea). Institute of Advanced Machinery Design; Lee, Kyu-Jung [Korea University, Seoul (Korea). Dept. of Mechanical Engineering

2005-08-01

Heat transfer and pressure drop characteristics of four microfin tubes were experimentally investigated for condensation of refrigerants R134a, R22, and R410A in four different test sections. The microfin tubes examined during this study consisted of 8.92, 6.46, 5.1, and 4 mm maximum inside diameter. The effect of mass flux, vapor quality, and refrigerants on condensation was investigated in terms of the heat transfer enhancement factor and the pressure drop penalty factor. The pressure drop penalty factor and the heat transfer enhancement factor showed a similar tendency for each tube at given vapor quality and mass flux. Based on the experimental data and the heat-momentum analogy, correlations for the condensation heat transfer coefficients in an annular flow regime and the frictional pressure drops are proposed. (author)

Compressed-tube pressure cell for optical studies at ocean pressures: Application to glucose mutarotation kinetics.

Science.gov (United States)

Lamelas, F J

2016-12-01

A self-contained compressed-tube pressure cell is tested to 25 MPa. The cell is very simple to construct and offers stable pressure control with optical access to fluid samples. The physical path length of light through the cell is large enough to measure optical activity. The entire system is relatively small and portable, and it is vibration-free, since a compressor is not used. Operation of the cell is demonstrated by measuring the mutarotation rate of aqueous glucose solutions at 25 °C. A logarithmic plot of the rate constant vs. pressure yields an activation volume for mutarotation of -22 cm 3 /mol, approximately twice the value measured previously at higher pressures.

Experimental investigation of pressure fluctuations caused by a vortex rope in a draft tube

International Nuclear Information System (INIS)

Kirschner, O; Ruprecht, A; Göde, E; Riedelbauch, S

2012-01-01

In the last years hydro power plants have taken the task of power-frequency control for the electrical grid. Therefore turbines in storage hydro power plants often operate outside their optimum. If Francis-turbines and pump-turbines operate at off-design conditions, a vortex rope in the draft tube can develop. The vortex rope can cause pressure oscillations. In addition to low frequencies caused by the rotation of the vortex rope and the harmonics of these frequencies, pressure fluctuations with higher frequencies can be observed in some operating points too. In this experimental investigation the flow structure and behavior of the vortex rope movement in the draft tube of a model pump-turbine are analyzed. The investigation focuses on the correlation of the pressure fluctuation frequency measured at the draft tube wall with the movement of the vortex rope. The movement of the vortex rope is analyzed by the velocity field in the draft tube which was measured with particle image velocimetry. Additionally, the vortex rope movement has been analyzed with the captures of high-speed-movies from the cavitating vortex rope. Besides the rotation of the vortex rope due to pressure fluctuation with low frequencies the results of the measurement also show a correlation between the rotation of the elliptical or deformed rope cross-section and the higher frequency pressure pulsation. An approximation shows that the frequencies of the pressure fluctuation and the movement of the vortex rope are also connected with the velocity of the flow. Taking into account the size and position of the cavitating vortex core as well as the velocity at the position of the surface of the cavitating vortex core the time-period of the rotation of the vortex core can be approximated. The results show that both, the low frequency pressure fluctuation and the higher frequency pressure fluctuation are correlating with the vortex rope movement. With this estimation, the period of the higher frequency

Non-destructive evaluation of stream generator tubes and pressure tubes from the PHWR reactors, using the rotating magnetic field method

International Nuclear Information System (INIS)

Premel, D.; Placko, D.; Grimberg, R.; Savin, A.

2001-01-01

This work presents a new type of eddy current transducer with a rotating magnetic field devoted to the inspection of steam generator tubes and pressure tubes from the PHWR reactors. A theoretical model has been developed that permits the calculations of the emf induced in the reception coils in the presence of the copper or magnetite deposits, anti-vibration railing and garter springs. (authors)

Replacement of a cracked pressure tube in Bruce GS unit 2

International Nuclear Information System (INIS)

Dunn, J.T.

1982-06-01

In 1982 February, a primary heat transport system leak was detected in the annulus gas system by on-line instrumentation. The source of the leak was found to be a small axial crack in the pressure tube of fuel channel X-14. This fuel channel was removed and replaced by station maintenance staff, and the unit was returned to service five weeks after it had been shut down. The cracked pressure tube was sent to Chalk River Nuclear Laboratories for examination, and the crack was found to be very similar to those found in Pickering GS units 3 and 4 in 1974-75. It was caused by delayed hydride cracking during the period of high residual stress between the time of rolling and the pre-service stress relief

Fabrication of Nb_3Al superconductor by the optimized mechanical alloying method with low temperature

International Nuclear Information System (INIS)

Zhang, Y.; Lin, W.J.; Xu, L.Y.; Yang, D.W.; Chen, Y.L.; Li, P.Y.; Pan, X.F.; Yan, G.; Zhao, Y.

2016-01-01

Highlights: • Due to a much better strain tolerance than Nb_3Sn, Nb_3Al has been considered as an excellent candidate for making high field magnets. At present, the Nb_3Al superconducting wires were prepared mainly by the Jelly-roll method combined with a rapid heating and quenching (RHQ) heat treatment at around 2000 °C. In this study, Nb_3Al superconductor with T_c of 15.6 K is directly prepared with a mechanical alloying method followed by a low temperature annealing at 800 to 900 °C. Our results hint the possibility that Nb_3Al superconducting wire with high performance can be prepared below the melting point of Cu (1080 °C) by a conventional powder in tube (PIT) method, thus effectively avoiding high temperature heat treatment and RHQ device. - Abstract: Mechanical alloying was used to synthesize Nb_3Al superconductor successfully, and the process was optimization under various preparation conditions. In the current study, Nb_3Al superconductor with T_c of 15.6 K was directly prepared from high quality Nb (Al) solid solution by mechanical alloying method and heat treatment at a low temperature of 800 to 900 °C. The results showed that Nb_3Al superconducting wire with high performance could be prepared after heat treatment below the melting point of Cu (1080°C) and using Nb (Al) solid solution and conventional powder in tube (PIT) method, thus effectively avoiding ultra-high temperature heat treatment and special rapid heating and quenching(RHQ) device.

Conceptual design of a pressure tube light water reactor with variable moderator control

International Nuclear Information System (INIS)

Rachamin, R.; Fridman, E.; Galperin, A.

2012-01-01

This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

A novel Fe–Cr–Nb matrix composite containing the TiB_2 neutron absorber synthesized by mechanical alloying and final hot isostatic pressing (HIP) in the Ti-tubing

International Nuclear Information System (INIS)

Litwa, Przemysław; Perkowski, Krzysztof; Zasada, Dariusz; Kobus, Izabela; Konopka, Gustaw; Czujko, Tomasz; Varin, Robert A.

2016-01-01

The Fe–Cr–Ti-Nb elemental powders were mechanically alloyed/ball milled with TiB_2 and a small quantity of Y_2O_3 ceramic to synthesize a novel Fe-based alloy-ceramic powder composite that could be processed by hot isostatic pressing (HIP) for a perceived potential application as a neutron absorber in nuclear reactors. After ball milling for the 30–80 h duration relatively uniform powders with micrometric sizes were produced. With increasing milling time a fraction of TiB_2 particles became covered with the much softer Fe-based alloy which resulted in the formation of a characteristic “core-mantel” structure. For the final HIP-ing process the mechanically alloyed powders were initially uniaxially pressed into rod-shaped compacts and then cold isostatically pressed (CIP-ed). Subsequently, the rod-shaped compacts were placed in the Ti-tubing and subjected to hot isostatic pressing (HIP) at 1150 °C/200 MPa pressure. The HIP-ing process resulted in the formation of the near-Ti and intermediate diffusional layers in the microstructure of HIP-ed samples which formed in accord with the Fe-Ti binary phase diagram. Those layers contain the phases such as α-Ti (HCP), the FeTi intermetallic and their hypo-eutectoid mixtures. In addition, needle-like particles were formed in both layers in accord with the Ti-B binary phase diagram. Nanohardness testing, using a Berkovich type diamond tip, shows that the nanohardness in the intermediate layer areas, corresponding to the composition of the hypo-eutectoid mixture of Ti-FeTi, equals 980.0 (±27.1) HV and correspondingly 1176.9 (±47.6) HV for the FeTi phase. The nanohardness in the sample's center in the areas with the fine mixture of Fe-based alloy and small TiB_2 particles equals 1048.3 (±201.8) HV. The average microhardness of samples HIP-ed from powders milled for 30 and 80 h is 588 HV and 733 HV, respectively. - Highlights: • A Fe–Cr–Nb-based composite with TiB_2 neutron absorbing ceramic was mechanically

Pressure Drop Correlations of Single-Phase and Two-Phase Flow in Rolling Tubes

International Nuclear Information System (INIS)

Xia-xin Cao; Chang-qi Yan; Pu-zhen Gao; Zhong-ning Sun

2006-01-01

A series of experimental studies of frictional pressure drop for single phase and two-phase bubble flow in smooth rolling tubes were carried out. The tube inside diameters were 15 mm, 25 mm and 34.5 mm respectively, the rolling angles of tubes could be set as 10 deg. and 20 deg., and the rolling periods could be set as 5 s, 10 s and 15 s. Combining with the analysis of single-phase water motion, it was found that the traditional correlations for calculating single-phase frictional coefficient were not suitable for the rolling condition. Based on the experimental data, a new correlation for calculating single-phase frictional coefficient under rolling condition was presented, and the calculations not only agreed well with the experimental data, but also could display the periodically dynamic characteristics of frictional coefficients. Applying the new correlation to homogeneous flow model, two-phase frictional pressure drop of bubble flow in rolling tubes could be calculated, the results showed that the relative error between calculation and experimental data was less than ± 25%. (authors)

Nb3Ge superconductive films grown with nitrogen

International Nuclear Information System (INIS)

Sigsbee, R.A.

1978-01-01

A superconductive film of Nb 3 Ge is produced by providing within a vacuum chamber a heated substrate and sources of niobium and germanium, reducing the pressure within the chamber to a residual pressure no greater than about 10 -5 mm Hg, introducing nitrogen into the resulting evacuated chamber in controlled amounts and vaporizing the niobium and germanium to deposit a film of Nb 3 Ge on the heated substrate

Response of the water status of soybean to changes in soil water potentials controlled by the water pressure in microporous tubes

Science.gov (United States)

Steinberg, S. L.; Henninger, D. L.

1997-01-01

Water transport through a microporous tube-soil-plant system was investigated by measuring the response of soil and plant water status to step change reductions in the water pressure within the tubes. Soybeans were germinated and grown in a porous ceramic 'soil' at a porous tube water pressure of -0.5 kpa for 28 d. During this time, the soil matric potential was nearly in equilibrium with tube water pressure. Water pressure in the porous tubes was then reduced to either -1.0, -1.5 or -2.0 kPa. Sap flow rates, leaf conductance and soil, root and leaf water potentials were measured before and after this change. A reduction in porous tube water pressure from -0.5 to -1.0 or -1.5 kPa did not result in any significant change in soil or plant water status. A reduction in porous tube water pressure to -2.0 kPa resulted in significant reductions in sap flow, leaf conductance, and soil, root and leaf water potentials. Hydraulic conductance, calculated as the transpiration rate/delta psi between two points in the water transport pathway, was used to analyse water transport through the tube-soil-plant continuum. At porous tube water pressures of -0.5 to-1.5 kPa soil moisture was readily available and hydraulic conductance of the plant limited water transport. At -2.0 kPa, hydraulic conductance of the bulk soil was the dominant factor in water movement.

The influence of the volume phase concentrations of the base material on the corrosion kinetics of Zr-2.5 Nb

International Nuclear Information System (INIS)

Olmedo, Ana M.; Bordoni, Roberto A.; Villegas, Marina; Hermida, Jorge D.

1999-01-01

The corrosion kinetics in lithiated heavy water at 350 and 265 C degrees of two unirradiated Zr-2.5 Nb pressure tube materials labelled C and J is presented. Both materials have the same behaviour up to 280 days of the exposure time and follow a para linear kinetic at 350 C degrees. At 265 C degrees, material C shows a larger weight gain and the formation of white oxide nuclei on the surface of the samples after 70 days of exposure. The size and coverage of oxide nuclei increase with the exposure time. X-ray diffraction analysis detected a difference of microstructure between the two materials, a larger volume content of the metastable β-Zr phase was found in C material. The lower β-Zr content of material J leads to a better corrosion behaviour than material C at the lower temperature (265 C degrees). At the higher temperature (350 C degrees), the decomposition Zr-β → ω+β Nb-enriq → α+ω+β Nb-enriq → α+β Nb-enriq → Zr-α + Nb-β induced at the corrosion test temperature occurs at a sufficiently fast rate so that no differences in corrosion behaviour are detected between both materials. (author)

Optimization of the heat treatment schedule for next european dipole (NED) powder in tube $Nb_{3}Sn$ strand

CERN Document Server

Boutboul, T; den Ouden, A; Pedrini, D; Volpini, G

2009-01-01

A Nb3Sn strand was successfully developed by the company SMI for Next European Dipole (NED) activity and on the basis of Powder-In-Tube (PIT) method. This strand, after the standard reaction recommended by the firm (84 h @ 675 oC), presents attractive performances as a critical current density in the non-copper part of ~ 2500 A/mm2 for 4.2 K and 12 T applied field, an effective filament diameter of ~ 50 μm and limited flux jumps at low magnetic fields. Heat treatment optimization studies are currently performed at CERN to try to optimize the strand electric abilities. For this purpose, various heat treatment schedules were already investigated with a plateau temperature as low as 625 oC. The preliminary results of these studies are summarized here.

A study on the critical heat flux for annuli and round tubes under low pressure conditions

International Nuclear Information System (INIS)

Park, Jae Wook

1997-02-01

This study aims to reveal the characteristics of the critical heat flux (CHF) of internally heated concentric annuli and vertical round tubes in low-pressure and low-flow (LPLF) conditions. Although many efforts have been devote to the subject of the CHF during the last forty years, the information on the CHF phenomenon for LPLF conditions is still very limited. The applicable ranges of the CHF correlations for annuli and round tubes are concentrate on the operating conditions of nuclear power plant (NPP), namely high-pressure and high-flow (HPHF) conditions. these facts promoted to collect the reliable CHF data for LPLF conditions for both annuli and round tubes. The critical heat flux data for vertical flow boiling of water in annuli and round tubes at low pressures and low mass fluxes show the following trends: The observed CHF mechanism for annuli was changed in the order of flooding, churn-to-annular flow transition, and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. The observed parametric trends for annuli are consistent with the previous understanding except that the CHF for downward flow is considerably lower (up to 40%) than that for upward flow. The critical quality is much lower than that for round tubes at the same inlet conditions. The observed parametric trends for round tubes are generally consistent with the previous understanding except for system pressure an tube diameter effect. For the system pressure effect, it is observed that the pressure effect is complicated but not so large, whereas the existing CHF correlations do not present the parametric trend exactly. For tube diameter effect, the decreasing trends of CHF with respect to tube diameter was the general understanding so far, but in this region the CHF show a increasing trend of tube diameter. The prediction and the parametric trend analyses are performed by two view points, I.e., for fixed inlet conditions and for local

Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Tsiklauri, G.; Schmitt, B.

1995-09-01

Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

Thermal-hydraulic instabilities in pressure tube graphite-moderated boiling water reactors

International Nuclear Information System (INIS)

Tsiklauri, G.; Schmitt, B.

1995-09-01

Thermally induced two-phase instabilities in non-uniformly heated boiling charmers in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement

Tube to tube excursive instability - sensitivities and transients

International Nuclear Information System (INIS)

Brown, M.; Layland, M.W.

1980-01-01

A simple basic analysis of excursive instability in a boiler tube shows how it depends upon operating conditions and physical properties. A detailed mathematical model of an AGR boiler is used to conduct a steady state parameter sensitivity survey. It is possible from this basis to anticipate the effects of changes in operating conditions and changes in design parameters upon tube to tube stability. Dynamic responses of tubes operating near the stability threshold are examined using a mathematical model. Simulated excursions are triggered by imparting small abrupt pressure changes on the boiler inlet pressure. The influences of the magnitude of the pressure change, waterside friction factor and gas side coupling between tubes are examined. (author)

Thermal performance and pressure drop of spiral-tube ground heat exchangers for ground-source heat pump

International Nuclear Information System (INIS)

Jalaluddin; Miyara, Akio

2015-01-01

Thermal performance and pressure drop of the spiral-tube GHE were evaluated in this present work. A numerical simulation tool was used to carry out this research. The heat exchange rates per meter borehole depth of the spiral-tube GHE with various pitches and their pressure drops were compared with that of the U-tube GHE. Furthermore, a comparative analysis between a spiral pipe and straight pipe was performed. In comparison with the straight pipe, using the spiral pipe in the borehole increased the heat exchange rate to the ground per meter borehole depth. However, the pressure drop of water flow also increased due to increasing the length of pipe per meter borehole depth and its spiral geometry. The accuracy of the numerical model was verified for its pressure drop with some pressure drop correlations. The heat exchange rate and pressure drop of the GHEs are presented. As an example, the heat exchange rate per meter borehole depth of spiral pipe with 0.05 m pitch in the turbulent flow increased of 1.5 times. Its pressure drop also increased of 6 times. However, from the view point of energy efficiency, using the spiral pipe in the ground-source heat pump system gives a better performance than using the straight pipe. The heat exchange rate and pressure drop are important parameter in design of the ground-source heat pump (GSHP) system. - Highlights: • Thermal performance and pressure drop of spiral-tube GHE are presented. • Effects of spiral pitch on thermal performance and pressure drop are analyzed. • Using a spiral pipe increases heat exchange rate per meter borehole depth of GHE. • Pressure drop per meter borehole depth also increases in the spiral pipe.