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Sample records for zirconium pressure tubes

  1. 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

  2. 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)

  3. 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

  4. 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.

  5. Multiaxial ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure

    Energy Technology Data Exchange (ETDEWEB)

    Chen, G.; Zhang, X. [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China); Xu, D.K. [Environmental Corrosion Center, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Li, D.H. [Hunan Taohuajiang Nuclear Power Co., Ltd, Yiyang, 413000 (China); Chen, X. [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China); Zhang, Z., E-mail: zhe.zhang@tju.edu.cn [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China)

    2017-06-15

    In this study, a series of uniaxial and multiaxial ratcheting tests were conducted at room temperature on zirconium alloy tubes. The experimental results showed that for uniaxial symmetrical cyclic test, the axial ratcheting strain ɛ{sub x} did not accumulate obviously in initial stage, but gradually increased up to 1% with increasing stress amplitude σ{sub xa}. For multiaxial ratcheting tests, the zirconium alloy tube was highly sensitive to both the axial stress amplitude σ{sub xa} and the internal pressure p{sub i}. The hoop ratcheting strain ɛ{sub θ} increased continuously with the increase of axial stress amplitude, whereas the evolution of axial ratcheting strain ɛ{sub x} was related to the axial stress amplitude. The internal pressure restricted the ratcheting accumulation in the axial direction, but promoted the hoop ratcheting strain on the contrary. The prior loading history greatly restrained the ratcheting behavior of subsequent cycling with a small internal pressure. - Highlights: •Uniaxial and multiaxial ratcheting behavior of the zirconium alloy tubes are investigated at room temperature. •The ratcheting depends greatly on the stress amplitude or internal pressure. •The interaction between the axial and hoop ratcheting mechanisms is greatly dependent on the internal pressure level. •The ratcheting is influenced significantly by the loading history of internal pressure.

  6. Multiaxial ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure

    International Nuclear Information System (INIS)

    Chen, G.; Zhang, X.; Xu, D.K.; Li, D.H.; Chen, X.; Zhang, Z.

    2017-01-01

    In this study, a series of uniaxial and multiaxial ratcheting tests were conducted at room temperature on zirconium alloy tubes. The experimental results showed that for uniaxial symmetrical cyclic test, the axial ratcheting strain ɛ x did not accumulate obviously in initial stage, but gradually increased up to 1% with increasing stress amplitude σ xa . For multiaxial ratcheting tests, the zirconium alloy tube was highly sensitive to both the axial stress amplitude σ xa and the internal pressure p i . The hoop ratcheting strain ɛ θ increased continuously with the increase of axial stress amplitude, whereas the evolution of axial ratcheting strain ɛ x was related to the axial stress amplitude. The internal pressure restricted the ratcheting accumulation in the axial direction, but promoted the hoop ratcheting strain on the contrary. The prior loading history greatly restrained the ratcheting behavior of subsequent cycling with a small internal pressure. - Highlights: •Uniaxial and multiaxial ratcheting behavior of the zirconium alloy tubes are investigated at room temperature. •The ratcheting depends greatly on the stress amplitude or internal pressure. •The interaction between the axial and hoop ratcheting mechanisms is greatly dependent on the internal pressure level. •The ratcheting is influenced significantly by the loading history of internal pressure.

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

    International Nuclear Information System (INIS)

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

    2009-02-01

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

  8. Automatic measuring system of zirconium thickness for zirconium liner cladding tubes

    International Nuclear Information System (INIS)

    Matsui, K.; Yamaguchi, H.; Hiroshima, T.; Sakamoto, T.; Murayama, R.

    1985-01-01

    An automatic system of pure zirconium liner thickness for zirconium-zircaloy cladding tubes has been successfully developed. The system consists of three parts. (1) An ultrasonic thickness measuring method for mother tubes before cold rolling. (2) An electromagnetic thickness measuring method for the manufactured tubes. (3) An image processing method for the cross sectional view of the manufactured cut tube samples. In Japanese nuclear industry, zirconium-zircaloy cladding tubes have been tested in order to realize load following operation in the atomic power plant. In order to provide for the practical use in the near future, Sumitomo Metal Industries, Ltd. has been studied and established the practical manufacturing process of the zirconium liner cladding tubes. The zirconium-liner cladding tube is a duplex tube comprising an inner layer of pure zirconium bonded to zircaloy metallurgically. The thickness of the pure zirconium is about 10 % of the total wall thickness. Several types of the automatic thickness measuring methods have been investigated instead of the usual microscopic viewing method in which the liner thickness is measured by the microscopic cross sectional view of the cut tube samples

  9. 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

  10. 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.)

  11. Localized deformation of zirconium-liner tube

    International Nuclear Information System (INIS)

    Nagase, Fumihisa; Uchida, Masaaki

    1988-03-01

    Zirconium-liner tube has come to be used in BWR. Zirconium liner mitigates the localized stress produced by the pellet-cladding interaction (PCI). In this study, simulating the ridging, stresses were applied to the inner surfaces of zirconium-liner tubes and Zircaloy-2 tubes, and, to investigate the mechanism and the extent of the effect, the behavior of zirconium liner was examined. As the result of examination, stress was concentrated especially at the edge of the deformed region, where zirconium liner was highly deformed. Even after high stress was applied, the deformation of Zircaloy part was small, since almost the concentrated stress was mitigated by the deformation of zirconium liner. In addition, stress and strain distributions in the cross section of specimen were calculated with a computer code FEMAXI-III. The results also showed that zirconium liner mitigated the localized stress in Zircaloy, although the affected zone was restricted to the region near the boundary between zirconium liner and Zircaloy. (author)

  12. 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

  13. 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.

  14. 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)

  15. Development of zirconium alloy tube manufacturing technology

    International Nuclear Information System (INIS)

    Kim, In Kyu; Park, Chan Hyun; Lee, Seung Hwan; Chung, Sun Kyo

    2009-01-01

    In late 2004, Korea Nuclear Fuel Company (KNF) launched a government funded joint development program with Westinghouse Electric Co. (WEC) to establish zirconium alloy tube manufacturing technology in Korea. Through this program, KNF and WEC have developed a state of the art facility to manufacture high quality nuclear tubes. KNF performed equipment qualification tests for each manufacturing machine with the support of WEC, and independently carried out product qualification tests for each tube product to be commercially produced. Apart from those tests, characterization test program consisting of specification test and characterization test was developed by KNF and WEC to demonstrate to customers of KNF the quality equivalency of products manufactured by KNF and WEC plants respectively. As part of establishment of performance evaluation technology for zirconium alloy tube in Korea, KNF carried out analyses of materials produced for the characterization test program using the most advanced techniques. Thanks to the accomplishment of the development of zirconium alloy tube manufacturing technology, KNF is expected to acquire positive spin off benefits in terms of technology and economy in the near future

  16. 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)

  17. 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)

  18. 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

  19. Delayed hydride cracking in Zr-2.5Nb pressure tubes

    International Nuclear Information System (INIS)

    Mieza, Juan I.; Domizzi, Gladys; Vigna, Gustavo L.

    2007-01-01

    Zr-2.5 Nb alloy from CANDU pressure tubes are prone to failure by hydrogen intake. One of the degradation mechanisms is delayed hydride cracking, which is characterized by the velocity of cracking. In this work, we study the effect of beta zirconium phase transformation over delayed hydride cracking velocity in Zr-2.5 Nb alloy from pressure tubes. Acoustic emission technique was used for cracking detection. (author) [es

  20. A computer model for hydride blister growth in zirconium alloys

    International Nuclear Information System (INIS)

    White, A.J.; Sawatzky, A.; Woo, C.H.

    1985-06-01

    The failure of a Zircaloy-2 pressure tube in the Pickering unit 2 reactor started at a series of zirconium hydride blisters on the outside of the pressure tube. These blisters resulted from the thermal diffusion of hydrogen to the cooler regions of the pressure tube. In this report the physics of thermal diffusion of hydrogen in zirconium is reviewed and a computer model for blister growth in two-dimensional Cartesian geometry is described. The model is used to show that the blister-growth rate in a two-phase zirconium/zirconium-hydride region does not depend on the initial hydrogen concentration nor on the hydrogen pick-up rate, and that for a fixed far-field temperature there is an optimum pressure-type/calandria-tube contact temperature for growing blisters. The model described here can also be used to study large-scale effects, such as hydrogen-depletion zones around hydride blisters

  1. 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

  2. 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

  3. Characteristics of Pilger Die Materials for Nuclear Zirconium Alloy Tubes

    International Nuclear Information System (INIS)

    Park, Ki Bum; Kim, In Kyu; Park, Min Young; Kahng, Jong Yeol; Kim, Sun Doo

    2011-01-01

    KEPCO Nuclear Fuel Company's (KEPCO NF) tube manufacturing facility, Techno Special Alloy (TSA) Plant, has started cold pilgering operation since 2008. It is obvious that the cold pilgering process is one of the key processes controlling the quality and the characteristics of the tubes manufactured, i.e. nuclear zirconium alloy tube in KEPCO NF. Cold pilgering is a rolling process for forming metal tubes in which diameter and wall thickness are reduced in a number of forming steps, using ring dies at outside of the tube and a curved mandrel at inside to reduce tube cross sections by up to 90 percent. The OD size of tube is reduced by a pair of dies, and ID size and wall thickness is controlled simultaneously by mandrel. During the cold pilgering process, both tools are the critical components for providing qualified tube. Development of pilger die and mandrel has been a significant importance in the zirconium tube manufacturing and a major goal of KEPCO NF. The objective of this study is to evaluate the life time of pilger die during pilgering. Therefore, a comparison of the heat treatment and mechanical properties of between AISI 52100 and AISI H13 materials was made in this study

  4. 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)

  5. Process for forming seamless tubing of zirconium or titanium alloys from welded precursors

    International Nuclear Information System (INIS)

    Sabol, G.P.; Barry, R.F.

    1987-01-01

    A process is described for forming seamless tubing of a material selected from zirconium, zirconium alloys, titanium, and titanium alloys, from welded precursor tubing of the material, having a heterogeneous structure resulting from the welding thereof. The process consists of: heating successive axial segments of the welded tubing, completely through the wall thereof, including the weld, to uniformly transform the heterogeneous, as welded, material into the beta phase; quenching the beta phase tubing segments, the heating and quenching effected sufficiently rapid enough to produce a fine sized beta grain structure completely throughout the precursor tubing, including the weld, and to prevent growth of beta grains within the material larger than 200 micrometers in diameter; and subsequently uniformly deforming the quenched precursor tubing by cold reduction steps to produce a seamless tubing of final size and shape

  6. Application of non-destructive liner thickness measurement technique for manufacturing and inspection process of zirconium lined cladding tube

    International Nuclear Information System (INIS)

    Nakazawa, Norio; Fukuda, Akihiro; Fujii, Noritsugu; Inoue, Koichi

    1986-01-01

    Recently, in order to meet the difference of electric power demand owing to electric power situation, large scale load following operation has become necessary. Therefore, the development of the cladding tubes which withstand power variation has been carried out, as the result, zirconium-lined zircaloy 2 cladding tubes have been developed. In order to reduce the sensitivity to stress corrosion cracking, these zirconium-lined cladding tubes require uniform liner thickness over the whole surface and whole length. Kobe Steel Ltd. developed the nondestructive liner thickness measuring technique based on ultrasonic flaw detection technique and eddy current flaw detection technique. These equipments were applied to the manufacturing and inspection processes of the zirconium-lined cladding tubes, and have demonstrated superiority in the control and assurance of the liner thickness of products. Zirconium-lined cladding tubes, the development of the measuring technique for guaranteeing the uniform liner thickness and the liner thickness control in the manufacturing and inspection processes are described. (Kako, I.)

  7. Polycrystalline models for the calculation of residual stresses in zirconium alloys tubes

    International Nuclear Information System (INIS)

    Signorelli, J.W.; Turner, P.A.; Lebensohn, R.A.; Pochettino, A.A.

    1995-01-01

    Tubes made of different Zirconium alloys are used in various types of reactors. The final texture of tubes as well as the distribution of residual stresses depend on the mechanical treatments done during their manufacturing process. The knowledge and prediction of both the final texture and the distribution of residual stresses in a tube for nuclear applications are of outstanding importance in relation with in-reactor performance of the tube, especially in what concerns to its irradiation creep and growth behaviour. The viscoplastic and the elastoplastic self consistent polycrystal models are used to investigate the influence of different mechanical treatments, performed during rolling processes on the final distribution of intergranular residual stresses of zirconium alloys tubes. The residual strains predictions with both formulations show a non linear dependence with the orientation, but they are qualitatively different. This discrepancy could be explain in terms of the relative plastic activity between the -type and -type deformation modes predicted with the viscoplastic and elastoplastic models. (author). 10 refs., 4 figs., 1 tab

  8. Tube in zirconium base alloy for nuclear fuel assembly and manufacturing process of such a tube

    International Nuclear Information System (INIS)

    Mardon, J.P.; Senevat, J.; Charquet, D.

    1996-01-01

    This patent concerns the description and manufacturing guidelines of a zirconium alloy tube for fuel cladding or fuel assembly guiding. The alloy contains (in weight) 0.4 to 0.6% of tin, 0.5 to 0.8% of iron, 0.35 to 0.50% of vanadium and 0.1 to 0.18% of oxygen. The carbon and silicon tenors range from 100 to 180 ppm and from 80 to 120 ppm, respectively. The alloy contains only zirconium, plus inevitable impurities, and is completely recrystallized. Corrosion resistance tests were performed on tubes made of this alloy and compared to corrosion tests performed on zircaloy 4 tubes. These tests show a better corrosion resistance and a lower corrosion kinetics for the new alloy, even in presence of lithium and iodine, and a lower hydridation rate. The mechanical resistance of this alloy is slightly lower than the one of zircaloy 4 but becomes equivalent or slightly better after two irradiation cycles. The ductility remains always equal or better than for zircaloy 4. (J.S.)

  9. 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%

  10. Analysis of hafnium in zirconium alloys

    International Nuclear Information System (INIS)

    Kondo, Isao; Sakai, Fumiaki; Ohuchi, Yoshifusa; Nakamura, Hisashi

    1977-01-01

    It is required to analyse alloying components and impurity elements in the acceptance analysis of zirconium alloys as the material for fuel cladding tubes and pressure tubes for advanced thermal reactors. Because of extreme similarity in chemical properties between zirconium and hafnium, about 100 ppm of hafnium is usually contained in zirconium alloys. Zircaloy-2 alloy and 2.5% Nb-zirconium with the addition of hafnium had been prepared as in-house standard samples for rapid analysis. Study was made on fluorescent X-ray analysis and emission spectral analysis to establish the analytical method. By using these in-house standard samples, acceptance analysis was successfully carried out for the fuel cladding tubes for advanced thermal reactors. Sulfuric acid solution was prepared from JAERI-Z 1, 2 and 3, the standard sample for zircaloy-2 prepared by the Analytical Committee on Nuclear Fuel and Reactor Materials, JAERI, and zirconium oxide (Hf 1 ppm/Zr). Standard Hf solution was added to the sulfuric acid solution step by step, to make up a series of the standard oxide samples by the precipitation process. By the use of these standard samples, the development of the analytical method and joint analysis were made by the three-member analytical technique research group including PNC. The analytical precision for the fluorescent X-ray analysis was improved by attaching a metallic yttrium filter to the window of an X-ray tube so as to suppress the effect due to zirconium matrix. The variation factor of the joint analysis was about 10% to show good agreement, and the indication value was determined. (Kobatake, H.)

  11. Zirconium - an imported mineral commodity

    International Nuclear Information System (INIS)

    1983-10-01

    This report examines Canada's position in regard to the principal zirconium materials: zircon; fusion-cast zirconium-bearing refractory products; zirconium-bearing chemicals; and zirconium metal, master alloys, and alloys. None of these is produced in Canada except fused alumina-zirconia and certain magnesium-zirconium alloys and zirconium-bearing steels. Most of the 3 000-4 000 tonnes of the various forms of zircon believed to be consumed in Canada each year is for foundry applications. Other minerals, notably chromite, olivine and silica sand are also used for these purposes and, if necessary, could be substituted for zircon. Zirconium's key role in Canada is in CANDU nuclear power reactors, where zirconium alloys are essential in the cladding for fuel bundles and in capital equipment such as pressure tubes, calandria tubes and reactivity control mechanisms. If zirconium alloys were to become unavailable, the Canadian nuclear power industry would collapse. As a contingency measure, Ontario Hydro maintains at least nine months' stocks of nuclear fuel bundles. Canada's vulnerability to short-term disruptions to supplies of nuclear fuel is diminished further by the availability of more expensive electricity from non-nuclear sources and, given time, from mothballed thermal plants. Zirconium minerals are present in many countries, notably Australia, the Republic of South Africa and the United States. Australia is Canada's principal source of zircon imports; South Africa is its sole source of baddeleyite. At this time, there are no shortages of either material. Canada has untapped zirconium resources in the Athabasca Oil Sands (zircon) and at Strange Lake along the ill-defined border between Quebec and Newfoundland (gittinsite). Adequate metal and alloy production facilities exist in France, Japan and the United States. No action by the federal government in regard to zirconium supplies is called for at this time

  12. 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)

  13. Effects of Oxidation and fractal surface roughness on the wettability and critical heat flux of glass-peened zirconium alloy tubes

    International Nuclear Information System (INIS)

    Fong, R.W.L.; Nitheanandan, T.; Bullock, C.D.; Slater, L.F.; McRae, G.A.

    2003-05-01

    Glass-bead peening the outside surfaces of zirconium alloy tubes has been shown to increase the Critical Heat Flux (CHF) in pool boiling of water. The CHF is found to correlate with the fractal roughness of the metal tube surfaces. In this study on the effect of oxidation on glass-peened surfaces, test measurements for CHF, surface wettability and roughness have been evaluated using various glass-peened and oxidized zirconium alloy tubes. The results show that oxidation changes the solid-liquid contact angle (i.e., decreases wettability of the metal-oxide surface), but does not change the fractal surface roughness, appreciably. Thus, oxidation of the glass-peened surfaces of zirconium alloy tubes is not expected to degrade the CHF enhancement obtained by glass-bead peening. (author)

  14. Determination of dislocation density in Zr-2.5Nb pressure tubes by x-ray

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Isaenkova, Perlovich; Cheong, Y. M.; Kim, S. S.; Yim, K. S.; Kwon, Sang Chul

    2000-11-01

    For X-ray determination of the dislocation density in CANDU Zr-2.5%Nb pressure tubes, a program was developed, using the Fourier analysis of X-ray line profiles and calculation of dislocation density by values of the coherent block size and the lattice distortion. The coincidence of obtained values of c- and a-dislocations with those, determined by the X-ray method for the same tube in AECL, was assumed to be the main criterion of validity of the developed program. The final variant of the program allowed to attain a rather close coincidence of calculated dislocation densities with results of AECL. The dislocation density was determined in all the zirconium grains with different orientations based on the texture of the stree-relieved CANDU tube. The complete distribution of c-dislocation density in -Zr grains depecding on their crystallographic orientations was constructed. The distribution of a-dislocation density within the texture maximum at L-direction, containing prismatic axes of all grains, was constructed as well. The analysis of obtained distributions testifies that -Zr grains of the stree-relieved CANDU tube significantly differ in their dislocation densities. Plotted diagrams of correlation between the dislocation density and the pole density allow to estimate the actual connection between texture and dislocation distribution in the studied tube. The distributions of volume fractions of all the zirconium grains depending on their dislocation density were calculated both for c- and a-dislocations. The distributions characterizes quantitatively the inhomogeneity of substructure conditions in the stress-relieved CANDU tube. the optimal procedure for determination of Nb content in {beta}-phases of CANDU Zr-2.5%Nb pressure tubes was also established.

  15. 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

  16. Growth and high pressure studies of zirconium sulphoselenide ...

    Indian Academy of Sciences (India)

    Growth and high pressure studies of zirconium sulphoselenide single ... tance was monitored in a Bridgman opposed anvil set-up up to 8 GPa pressure to identify .... The optical band gaps of the as-grown crystals were obtained by optical ab-.

  17. 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)

  18. Fretting wear behavior of zirconium alloy in B-Li water at 300 °C

    Science.gov (United States)

    Zhang, Lefu; Lai, Ping; Liu, Qingdong; Zeng, Qifeng; Lu, Junqiang; Guo, Xianglong

    2018-02-01

    The tangential fretting wear of three kinds of zirconium alloys tube mated with 304 stainless steel (SS) plate was investigated. The tests were conducted in an autoclave containing 300 °C pressurized B-Li water for tube-on-plate contact configuration. The worn surfaces were examined with scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS) and 3D microscopy. The cross-section of wear scar was examined with transmission electron microscope (TEM). The results indicated that the dominant wear mechanism of zirconium alloys in this test condition was delamination and oxidation. The oxide layer on the fretted area consists of outer oxide layer composed of iron oxide and zirconium oxide and inner oxide layer composed of zirconium oxide.

  19. 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)

  20. Development of microstructure in thermomechanical processing of zirconium alloys

    International Nuclear Information System (INIS)

    Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

    2009-01-01

    Zirconium based alloys are used for the manufacture of fuel tubes pressure tubes calandria tubes and other components of Pressurized Heavy Water Reactors (PHWRS). In single or two phase zirconium alloy system a variety of microstructure can be generated by suitable heat treatments by the process of equilibrium and non equilibrium phase transformations Microstructure can also be modified by alloying with α and β stabilizers. The microstructure in Zr alloys could be single hexagonal phase (α alloys) two phase bcc and hexagonal (α + β alloys) phase, single metastable martensitic microstructure and β with ω phase. The microstructural and micro textural evolution during thermo mechanical treatments depends strongly on such initial microstructure. Hot extrusion is a significant bulk deformation step which decides the initial microstructure of the alloy. It is carried out at elevated temperature i e above the recrystallization temperature, which enable imposition of large strains in single step. This deformation causes a significant change in the microstructure of the material and depends on extrusion process parameters such as temperature, strain rate (Ram speed), reduction ratio etc. In the present paper development of microstructures, microtexture and texture have been examined. An attempt is also made to optimise the hot working parameters for different Zirconium alloys with help of these studies. (author)

  1. Hot hardness studies on zircaloy 2 pressure tube along three orientations

    International Nuclear Information System (INIS)

    Kutty, T.R.G.; Ravi, K.; Jarvis, T.; Sengupta, A.K.; Majumdar, S.; Tewari, R.; Shrivastava, D.; Dey, G.K.

    2002-01-01

    Zirconium based alloys are the natural choice for both the fuel element cans and in-core structural components in water cooled nuclear reactors. In this paper, the hot hardness behaviour of zircaloy 2 pressure tubes has been examined from room temperature to 400 degC using a hot hardness tester. For the purpose of comparison, the hardness of the as cast and room temperature rolled specimens has also been carried out. For this, the samples were cut along three orientations and hardness was measured in each of these directions using Vickers diamond pyramid indenter. The variation in hardness of the pressure tube samples show that the hardness was highest along circumferential direction and least along the axial direction. The room temperature rolled samples showed highest hardness along the rolling planes. These variations in hardness could be explained in terms of development of texture during working on the material. (author)

  2. 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.)

  3. 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.

  4. 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

  5. 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

  6. Zirconium alloy fuel cladding resistant to PCI crack propagation

    International Nuclear Information System (INIS)

    Boyle, R.F.; Foster, J.P.

    1987-01-01

    A nuclear fuel element is described cladding tube comprising: concentric tubular layers of zirconium base alloys; the concentric tubular layers including an inner layer and outer layer; the outer layer metallurgically bonded to the inner layer; the outer layer composed of a first zirconium base alloy characterized by excellent resistance to corrosion caused by exposure to high temperature and pressure aqueous environments; the inner layer composed of a second zirconium base alloy consisting of: about 0.2 to 0.6 wt.% tin, about 0.03 to 0.11 wt.% iron, less than about 0.02 wt.% chromium, up to about 350 ppm oxygen and the remainder being zirconium and incidental impurities, and the inner layer characterized by improved resistance to crack propagation under reactor operating conditions compared to the first zirconium alloy

  7. 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.)

  8. Electron microscopy of nuclear zirconium alloys

    International Nuclear Information System (INIS)

    Versaci, R.A.; Ipohorski, Miguel

    1986-01-01

    Transmission electron microscopy observations of the microstructure of zirconium alloys used in fuel sheaths of nuclear power reactors are reported. Specimens were observed after different thermal and mechanical treatment, similar to those actually used during fabrication of the sheaths. Electron micrographs and electron diffraction patterns of second phase particles present in zircaloy-2 and zircaloy-4 were also obtained, as well as some characteristic parameters. Images of oxides and hydrides most commonly present in zirconium alloys are also shown. Finally, the structure of a Zr-2,5Nb alloy used in CANDU reactors pressure tubes, is observed by electron microscopy. (Author) [es

  9. 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.)

  10. Fuel assembly and fuel cladding tube

    International Nuclear Information System (INIS)

    Tsutsumi, Shinro; Ito, Ken-ichi; Inagaki, Masatoshi; Nakajima, Junjiro.

    1996-01-01

    A fuel cladding tube is a zirconium liner tube formed by lining a pure zirconium layer on the inner side of a zirconium alloy tube. The fuel cladding tube is formed by extrusion molding of a composite billet formed by inserting a pure zirconium billet into a zirconium alloy billet. Accordingly, the pure zirconium layer and the zirconium alloy tube are strongly joined by metal bond. The fuel cladding tube has an external oxide film on the outer surface of the zirconium alloy tube and an internal oxide film on the inner side of the pure zirconium layer. The external oxide film has a thickness preferably of about 1μm. The internal oxide film has a thickness of not more than 10μm, preferably, from 1 to 5μm. With such a constitution, flaws to be formed on both inner and outer surfaces of the cladding tube upon assembling a fuel assembly can be reduced thereby enabling to reduce the amount of hydrogen absorbed to the cladding tube. (I.N.)

  11. 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)

  12. Pressure of saturated vapor of yttrium and zirconium acetylacetonates

    Energy Technology Data Exchange (ETDEWEB)

    Trembovetskij, G.V.; Berdonosov, S.S.; Murav' eva, I.A.; Martynenko, L.I. (Moskovskij Gosudarstvennyj Univ. (USSR))

    1984-08-01

    The static method and the flow method using /sup 91/Y and /sup 95/Zr radioactive indicators have been applied to determine pressure of saturated vapour of yttrium and zirconium acetylacetonates. Values of thermodynamic functions ..delta..Hsub(subl)=(98+-16)kJ/mol and ..delta..Ssub(subl.)=(155+-30)J/mol x K are calculated for sublimation of yttrium acetylacetonate. For sublimation of zirconium acetylacetonates ..delta..Hsub(subl) equals (116+-38) kJ/mol and ..delta..Ssub(subl) is equal to (198+-65) J/molxK.

  13. Modification of structural phase state in superficial layers of fuel tubes made of Zirconium alloys

    International Nuclear Information System (INIS)

    Volkov, N.; Kalin, B.; Pimenov, Y.; Timoshin, S.

    2011-01-01

    The paper presents the results obtained in developing the method for introduction of the required changes into states and properties of outer surface on fuel rod cladding made of zirconium alloys E110 and E635 through irradiation by radial Ar + ion beam with a broad energy spectrum. In particular, the paper demonstrates that ion beam treatment of the claddings surface, at the final stage of their fabrication, can upgrade substantially quality of outer tubular surface after mechanical polishing (the cleaner surface, the lower roughness, removal of technological transversal scratches). In addition, the ion beam irradiation results in higher micro-hardness of the modified layer and in better tribological parameters. Kinetic effects in growth of oxide films were studied for the tubular samples of zirconium alloys after ion-beam treatment (cleaning and polishing by radial Ar + ion beam). Also, corrosion tests of the tubular samples were carried out in water (at 350 0 C) and steam (at 350, 375 and 400 0 C) with duration up to 3000 hours. It was revealed that oxide layer consisting mainly of zirconium dioxide in monoclinic modification was formed on tubular surface after oxidation at 3500 0 C in water or steam. The oxidizing process in the pressurized steam created thicker oxide layer on tubular surface than that in the pressurized water. Experimental data were used to determine optimal conditions for ion-beam treatment of outer fuel tube surface. The tubular samples with the following geometrical parameters were investigated: length - up to 500 mm, diameter - 9,15 mm. Optimal regimes for ion-beam cleaning and polishing of the tubular samples were studied up to the process rate of 1 meter per minute. Within the frames of linear approximation, analytical relationships were derived for time dependent growth of oxide films and used to evaluate thickness of oxide film under test conditions (duration . up to 10000 hours). Thickness of oxide films can cover the range from 6

  14. Design basis for creep of zirconium alloy components in a fast neutron flux

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Fidleris, V.

    1975-01-01

    The chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behaviour of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

  15. Design basis for creep of zirconium alloy components in a fast neutron flux

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Fidleris, V.

    1974-01-01

    The Chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behavior of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time, and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

  16. Prospects for zirconium structural alloys at high temperatures

    International Nuclear Information System (INIS)

    Thomas, W.R.

    1969-05-01

    Improved station efficiencies and lower capital costs provide incentives for the development of zirconium alloys for pressure tubes which can operate at temperatures above 450 o C. The experience of the Ti industry indicates that a complex alloy containing solution hardeners of Sn or Al and precipitation hardeners of Mo and Nb and perhaps Si will be required. The thermal neutron cross-section of the alloy will be about 10% higher than Zircaloy-2 and because of its poor corrosion resistance will require cladding with a corrosion resistant alloy such as Zr-Cr. Results to date indicate that such a pressure tube is feasible. (author)

  17. 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)

  18. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes. Determination of hydrogen concentration and blister characterization

    International Nuclear Information System (INIS)

    2009-03-01

    Heavy water reactors (HWRs) comprise significant numbers of today's operating nuclear power plants, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes, are an important factor in ensuring 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 Intercomparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the framework of the IAEA's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of IAEA's project on advanced technologies for HWRs. The objective of the CRP was to compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP participants investigated the capability of different techniques to detect and characterize flaws. During the second phase participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in zirconium alloys. The intention was to identify the most effective pressure tube inspection and diagnostic methods and to identify further development needs. The organizations which participated in phase 2 of this CRP are: - Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL), Chalk River Laboratories (CRL), Canada; - Bhabha Atomic Research Centre (BARC), India; - Korea Atomic Energy Research Institute (KAERI), Republic of Korea; - National Institute for Research and Development for Technical Physics (NIRDTP), Romania; - Nuclear Non-Destructive Testing Research and Services (NNDT), Romania. IAEA-TECDOC-1499

  19. Extraction of zirconium from raffinate stream of Zirconium Oxide Plant raffinate

    International Nuclear Information System (INIS)

    Pandey, Garima; Chinchale, R.; Renjith, A.U.; Mukhopadhyay, S.; Shenoy, K.T.; Ghosh, S.K.

    2013-01-01

    Solution. The process conditions were, phase ratio for dispersion O/A: 1, organic: 0.1 M MAPO in 80:20 dodecane: isodecanol, strip: 0.2 M Oxalic acid and the agitation speed for dispersion: 3000 RPM. Feed solution at tube side was kept at a pressure of 1.5 Psig higher than the dispersion at shell-side. Flowrate of feed solution from shell side was maintained at 200 mlpm and that of dispersion at tube side was kept at 600 mlpm. The extent of extraction achieved after 4 hours of recirculation was 69%. Strip concentration of zirconium obtained was 0.628 g/L leading to 81 % stripping. The studies show that with adequate optimization of operating parameters, the HFDLM process shows promise for recovery of the Zr from lean streams like raffinate of Zr plant

  20. Nuclear power - replacement of pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The CANDU pressure tube reactor is an effective electricity generator. While most units have been built in Canada, units are successfully operated in Argentina and Korea as well as India and Pakistan, which have early versions of the same concept. Units are also under construction in Korea and Romania. The main constructional components of a CANDU core are the calandria vessel, the fuel channels and the reactivity control mechanisms. The fuel channel, in particular the pressure tubes, see an environment comprising high flux, high temperature water at high pressures, which induces changes in the properties and dimensions of the channel components. From the first, fuel channels were designed to be replaced because of the difficulty in predicting the behaviour of zirconium alloys in such service over a long period of time. In fact some phenomena, that were not known at the time of the earliest designs, have led to unacceptable changes in the properties of the channels and these early reactors have had to be retubed at half their intended life. These deficiencies have been corrected in the latest designs and fuel channels in reactors that have commenced operation over the last 10 years, are predicted to reach the intended 30 years life before replacement is necessary. The changing of fuel channels, the details and experience of which are explained, has been shown to be an effective way of refurbishing the CANDU reactor, extending its lifetime a further 25-30 years. (author)

  1. 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.)

  2. X-ray study of texture in zirconium alloy tubes and in graphite

    International Nuclear Information System (INIS)

    Skvortsov, V.V.; Alekseev, S.I.

    1987-01-01

    X-ray study of texture in zirconium alloy tubes and in graphite has been developed. The method is based on constructing coordinate grid of stereographic projection determining quantity and coordinates of points where measurements should be performed depending on a specimen slope pitch. Complete stereographic projection obtained so is a base both for constructing pole figures showing distribution normales of plane system being studied and for calculating texture coefficients determining property anisotropy in materials under investigation. This method can be applied to study texture in items of any materials independent of the item shape

  3. Determination of delayed hydride cracking velocity of CANDU Zr-2.5Nb pressure tube

    International Nuclear Information System (INIS)

    Kim, Young Suk; Kim, Chan Jung; Rheem, Y. W.; Im, K. S.; Kwon, Sang Chul

    2000-07-01

    As agreed upon the contract with an IAEA Co-ordinated Research Project 'Hydrogen and Hydride Induced Degradation of the Mechanical and Physical Properties of Zirconium Based Alloys', we conducted DHC tests at 3 different temperatures of 144, 182 and 250 deg C on the curved compact tension specimens made from a Zr-2.5Nb pressure tube. Additional tests were carried out at 200 and 230 deg C with an aim to determine the activation energy for delayed hydride cracking. This report summarizes the results of DHC tests obtained so far. All the DHC tests were conducted in accordance with the procedures suggested by the Host Lab. 7 DHCV values determined at the same temperature such as 250 deg C show very low standard deviation, whose average values are very comparable to those reported by the participants. Thus, one of the most important results we have got is that we establish qualified DHC testing procedure through the IAEA CRP. An activation energy for DHC of unirradiated Zr-2.5Nb pressure tube was 49 KJ/mol which is very similar to the activation energy of 43 KJ/mol for irradiated Zr-2.5Nb pressure tubes. DHCV increased linearly with the hydrogen content up to around 25 ppm and then became saturated at higher hydrogen concentration

  4. 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.)

  5. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes: Flaw detection and characterization [Phase 1] [Sample summary reports of pressure tube samples from Argentina, India, Canada, Republic of Korea, and Romania

    International Nuclear Information System (INIS)

    2006-05-01

    Nuclear power plants with heavy water reactors (HWRs) comprise nine percent of today's operating nuclear units, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems 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 Programme (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the frame of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of the IAEA's project on advanced technologies for HWRs. The objective of the CRP was to inter-compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP, participants have investigated the capability of different techniques to detect and characterize flaws. During the second phase of this CRP, participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in Zirconium alloys. The intent was to identify the most effective pressure tube inspection and diagnostic methods, and to identify further development needs. The organizations that have participated in this CRP are: - The Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL); Chalk River Laboratories (CRL), Canada; - The Research Institute of Nuclear Power Operations (RINPO), China National Nuclear Corporation (CNNC), China; - Bhabha Atomic Research Centre (BARC), India; - The Korea Electric Power Research Institute (KEPRI), Republic of Korea; - The Korea Atomic Energy

  6. 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.

  7. 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

  8. Recovery of zirconium from pickling solution, regeneration and its reuse

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharjee, D. [Nuclear Fuel Complex, Hyderabad 500062 (India); Mandal, D., E-mail: dmandal10@gmail.com [Alkali Material & Metal Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Visweswara Rao, R.V.R.L.; Sairam, S.; Thakur, S. [Nuclear Fuel Complex, Hyderabad 500062 (India)

    2017-05-15

    Graphical abstract: The following compares the performance of fresh pickling solution (PS) and regenerated and used pickling solution (UPS). - Highlights: • Pickling of zircaloy tubes and appendages is carried out to remove oxide layer. • The pickling solution become saturated with zirconium due to reuse. • As NaNO{sub 3} concentration increases, conc. of Zr in pickling solution decreases. • Experimental results shows that, used pickling solution can be regenerated. • Regenerated solution may be reused by adding makeup quantities of HF-HNO{sub 3}. - Abstract: The pressurized heavy water reactors use natural uranium oxide (UO{sub 2}) as fuel and uses cladding material made up of zircaloy, an alloy of zirconium. Pickling of zircaloy tubes and appendages viz., spacer and bearing pads is carried out to remove the oxide layer and surface contaminants, if present. Pickling solution, after use for many cycles i.e., used pickling solution (UPS) is sold out to vendors, basically for its zirconium value. UPS, containing a relatively small concentration of hydrofluoric acid. After repeated use, pickling solution become saturated with zirconium fluoride complex and is treated by adding sodium nitrate to precipitate sodium hexafluro-zirconate. The remaining solution can be recycled after suitable makeup for further pickling use. The revenue lost by selling UPS is very high compared to its zirconium value, which causes monetary loss to the processing unit. Experiments were conducted to regenerate and reuse UPS which will save a good amount of revenue and also protect the environment. Experimental details and results are discussed in this paper.

  9. Expanded heat treatment to form residual compressive hoop stress on inner surface of zirconium alloy tubing

    International Nuclear Information System (INIS)

    Megata, Masao

    1997-01-01

    A specific heat treatment process that introduces hoop stress has been developed. This technique can produce zirconium alloy tubing with a residual compressive hoop stress near the inner surface by taking advantage of the mechanical anisotropy in hexagonal close-packed zirconium crystal. Since a crystal having its basal pole parallel to the tangential direction of the tubing is easier to exhibit plastic elongation under the hoop stress than that having its basal pole parallel to the radial direction, the plastic and elastic elongation can coexist under a certain set of temperature and hoop stress conditions. The mechanical anisotropy plays a role to extend the coexistent stress range. Thus, residual compressive hoop stress is formed at the inner surface where more plastic elongation occurs during the heat treatment. This process is referred to as expanded heat treatment. Since this is a fundamental crystallographic principle, it has various applications. The application to improve PCI/SCC (pellet cladding interaction/stress corrosion cracking) properties of water reactor fuel cladding is promising. Excellent results were obtained with laboratory-scale heat treatment and an out-reactor iodine SCC test. These results included an extension of the time to SCC failure. (author)

  10. 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

  11. 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

  12. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

    2008-01-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

  13. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    Science.gov (United States)

    Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

    2008-12-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

  14. Tensile properties of quadruple melted Zr-2.5Nb pressure tubes evaluated from pressure tube offcuts

    International Nuclear Information System (INIS)

    Shah, Priti Kotak; Dubey, J.S.; Anantharaman, S.

    2013-12-01

    Rajasthan Atomic Power Station-2 (RAPS-2) is the first Pressurised Heavy Water Reactor (PHWR) in India having quadruple melted Zr-2.5Nb pressure tubes. Front-end and back-end off-cuts of sixteen pressure tubes were selected for studying the mechanical properties in axial and transverse directions of the tube. Tension tests were carried out at room temperature and at 300℃ using miniature tensile test specimens. The report presents the experimental details and discusses the base line tensile property data for the quadruple melted pressure tubes of RAPS-2. This data will be useful for the reactor life management. (author)

  15. 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.

  16. Hydrogen charging, hydrogen content analysis and metallographic examination of hydride in zirconium alloys

    International Nuclear Information System (INIS)

    Singh, R.N.; Kishore, R.; Mukherjee, S.; Roychowdhury, S.; Srivastava, D.; Sinha, T.K.; De, P.K.; Banerjee, S.; Gopalan, B.; Kameswaran, R.; Sheelvantra, Smita S.

    2003-12-01

    Gaseous and electrolytic hydrogen charging techniques for introducing controlled amount of hydrogen in zirconium alloy is described. Zr-1wt%Nb fuel tube, zircaloy-2 pressure tube and Zr-2.5Nb pressure tube samples were charged with up to 1000 ppm of hydrogen by weight using one of the aforementioned methods. These hydrogen charged Zr-alloy samples were analyzed for estimating the total hydrogen content using inert gas fusion technique. Influence of sample surface preparation on the estimated hydrogen content is also discussed. In zirconium alloys, hydrogen in excess of the terminal solid solubility precipitates out as brittle hydride phase, which acquire platelet shaped morphology due to its accommodation in the matrix and can make the host matrix brittle. The F N number, which represents susceptibility of Zr-alloy tubes to hydride embrittlement was measured from the metallographs. The volume fraction of the hydride phase, platelet size, distribution, interplatelet spacing and orientation were examined metallographically using samples sliced along the radial-axial and radial-circumferential plane of the tubes. It was observed that hydride platelet length increases with increase in hydrogen content. Considering the metallographs generated by Materials Science Division as standard, metallographs prepared by the IAEA round robin participants for different hydrogen concentration was compared. It is felt that hydride micrographs can be used to estimate not only that approximate hydrogen concentration of the sample but also its size, distribution and orientation which significantly affect the susceptibility to hydride embrittlement of these alloys. (author)

  17. 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)

  18. 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

  19. 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.

  20. 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

  1. High pressure low temperature hot pressing method for producing a zirconium carbide ceramic

    Science.gov (United States)

    Cockeram, Brian V.

    2017-01-10

    A method for producing monolithic Zirconium Carbide (ZrC) is described. The method includes raising a pressure applied to a ZrC powder until a final pressure of greater than 40 MPa is reached; and raising a temperature of the ZrC powder until a final temperature of less than 2200.degree. C. is reached.

  2. An experimental investigation of the rate of hydrogen absorption in zirconium-2.5 wt percent niobium from steam/hydrogen mixtures at elevated temperatures

    International Nuclear Information System (INIS)

    Langman, V.J.

    1984-08-01

    The test specifications for an experimental program to study the rate of hydrogen absorption in zirconium-2.5 weight percent niobium pressure tube material from steam/hydrogen mixtures at elevated temperatures are discussed

  3. 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

  4. 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

  5. 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.)

  6. Neutronographic Texture Analysis of Zirconium Based Alloys

    International Nuclear Information System (INIS)

    Kruz'elová, M; Vratislav, S; Kalvoda, L; Dlouhá, M

    2012-01-01

    Neutron diffraction is a very powerful tool in texture analysis of zirconium based alloys used in nuclear technique. Textures of five samples (two rolled sheets and three tubes) were investigated by using basal pole figures, inversion pole figures, and ODF distribution function. The texture measurement was performed at diffractometer KSN2 on the Laboratory of Neutron Diffraction, Department of Solid State Engineering, Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. Procedures for studying textures with thermal neutrons and procedures for obtaining texture parameters (direct and inverse pole figures, three dimensional orientation distribution function) are also described. Observed data were processed by software packages HEXAL and GSAS. Our results can be summarized as follows: i) All samples of zirconium alloys show the distribution of middle area into two maxima in basal pole figures. This is caused by alloying elements. A characteristic split of the basal pole maxima tilted from the normal direction toward the transverse direction can be observed for all samples, ii) Sheet samples prefer orientation of planes (100) and (110) perpendicular to rolling direction and orientation of planes (002) perpendicular to normal direction, iii) Basal planes of tubes are oriented parallel to tube axis, meanwhile (100) planes are oriented perpendicular to tube axis. Level of resulting texture and maxima position is different for tubes and for sheets. The obtained results are characteristic for zirconium based alloys.

  7. 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

  8. Separation of hafnium from zirconium in sulfuric acid solutions using pressurized ion exchange

    International Nuclear Information System (INIS)

    Hurst, F.J.

    1981-01-01

    High-resolution pressurized ion exchange has been used successfully to study and separate hafnium and zirconium sulfate complexes by chromatographic elution from Dowex 50W-X8 (15 to 25 μm) resin with sulfuric acid solutions. Techniques were developed to continuously monitor the column effluents for zirconium and hafnium by reaction with fluorometric and colorimetric reagents. Since neither reagent was specific for either metal ion, peak patterns were initially identified by using the stable isotopes 90 Zr and 180 Hf as fingerprints of their elution position. Distribution ratios for both zirconium and hafnium decrease as the inverse fourth power of the sulfuric acid concentration below 2N and as the inverse second power at higher acid concentration. The hafnium-to-zirconium separation factor is approximately constant (approx. 8) over the 0.5 to 3N range. Under certain conditions, an unseparated fraction was observed that was not retained by the resin. The amount of this fraction which is thought to be a polymeric hydrolysis product appears to be a function of metal and sulfuric acid concentrations. Conditions are being sought to give the highest zirconium concentration and the lowest acid concentration that can be used as a feed material for commercial scale-up in the continuous annular chromatographic (CAC) unit without formation of the polymer

  9. 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...

  10. Method of processing spent fuel cladding tubes

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi; Ouchi, Atsuhiro; Imahashi, Hiromichi.

    1986-01-01

    Purpose: To decrease the residual activity of spent fuel cladding tubes in a short period of time and enable safety storage with simple storage equipments. Constitution: Spent fuel cladding tubes made of zirconium alloys discharged from a nuclear fuel reprocessing step are exposed to a grain boundary embrittling atmosphere to cause grain boundary destruction. This causes grain boundary fractures to the zirconium crystal grains as the matrix of nuclear fuels and then precipitation products precipitated to the grain boundary fractures are removed. The zirconium constituting the nuclear fuel cladding tube and other ingredient elements contained in the precipitation products are separated in this removing step and they are separately stored respectively. As a result, zirconium constituting most part of the composition of the spent nuclear fuel cladding tubes can be stored safely at a low activity level. (Takahashi, M.)

  11. 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)

  12. 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.)

  13. Failure maps for internally pressurized Zr-2.5% Nb pressure tubes with circumferential temperature variations

    International Nuclear Information System (INIS)

    Shewfelt, R.S.W.

    1986-01-01

    During some postulated loss-of-coolant accidents, the pressure tube temperature may rise before the internal pressure drops, causing the pressure tube to balloon. The temperature around the pressure tube circumference would likely be nonuniform, producing localized deformation that could possibly cause failure. The computer program, GRAD, was used to determine the circumferential temperature distribution required to cause an internally pressurized Zr-2.5% Nb pressure tube to fail before coming into full contact with its calandria tube. These results were used to construct failure maps. 7 refs

  14. Development of heat treated Zr-2.5% Nb alloy tubes for pressure tubes

    International Nuclear Information System (INIS)

    Saibaba, N.; Jha, S.K.; Tonpe, S.

    2011-01-01

    Zr-2.5% Nb alloy is the candidate material for pressure tubes of Pressurized Heavy Water Reactors (PHWR), and are manufactured in cold working condition while heat treated pressure tubes are used in RBMK and FUGEN type of reactors. The diametral creep of these tubes is the life limiting factor. This paper presents the extensive work carried out for the optimization of process parameters to manufacture heat treated Zr-2.5% Nb pressure tubes. Extensive dilactometry study was carried out to establish the transus temperature for the alloy and the effect of soaking temperature and cooling rate on the microstructure was characterized. On the basis of the study, water quenching (at 883 deg C) in the a b region with 20-25% primary a phase was selected, further cold worked, aged and finally autoclaved. Mechanical properties of the finished tubes were found to be comparable to the cold worked route. Large number of full sized tubes of about 700 - 800 mm long was produced to establish the repeatability. (author)

  15. The development of zirconium alloy and its manufacturing

    International Nuclear Information System (INIS)

    Yuan Gaihuan; Yue Qiang

    2015-01-01

    Nuclear power which acts as one of low-carbon energy resources is the most realistic in large-scale application. It is also the preferred choice for many countries to develop energy resources and optimize its structure. Zirconium alloy is a key structural material for nuclear power plant fuel assemblies and cladding tubes of zirconium alloy are often referred as the first safeguard to nuclear power safety. With the development of nuclear power, three kinds of zirconium alloys Zr-Sn, Zr-Nb, Zr-Sn-Nb and with the representative products of Zr-4, M5, Zirlo respectively are developed and widely applied. Because of its severe operating environment and influence to nuclear safety, the requirements to zirconium alloys for physical and chemical properties, nuclear capability, tolerance and surface quality are very strict. The in-depth research and its manufacture capability become one of the main barriers for many countries who are developing the nuclear energy. In recent years, a stated-owned company, State Nuclear Bao Ti Zirconium Industry Company ('SNZ' for short) as well as National R and D Center for Nuclear Grade Zirconium material, is founded to meet the requirement of the rapid development of China's nuclear power industry. SNZ is dedicated for the fabrication and the research of nuclear grade zirconium products. After the successful completion of technology transfer of manufacturing for production chain and fully grasped of the manufacturing technology for the nuclear grade zirconium sponge through zirconium alloy tube, rod and strip products. National R and D Center for Nuclear Grade Zirconium material is cooperating with universities, nuclear energy research and design institutes and the owners of nuclear power plant to develop new zirconium alloy of self-owned brand. Through the selection of components, in-process testing and product inspection, four kinds of new zirconium alloys owns better performance than currently commercialized M5, Zirlo etc

  16. 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.

  17. 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

  18. 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)

  19. Thermal creep behavior of N36 zirconium alloy cladding tube

    International Nuclear Information System (INIS)

    Wang, P.; Zhao, W.; Dai, X.

    2015-01-01

    N36 is an alloy containing Zr, Sn, Nb and Fe that is developed by China as a superior cladding material to meet the performance of PWR fuel assembly at the maximum fuel rod burn-up. The creep characteristics of N36 zirconium alloy cladding tube were investigated at temperature from 593 K to 723 K with stress ranging from 20 MPa to 160 MPa. Transitions in creep mechanisms were noted, showing the distinct three rate-controlled creep mechanisms for the alloy at test conditions. In the region of low stresses with stress exponent n ∼ 1 and activation energy Q ∼ (104±4) kJ.mol -1 , Coble creep, based on diffusion of materials through grain boundaries, is the dominant rate-controlling mechanism, which contributes to the creep deformation. The formation of slip bands acts as an accommodation mechanism. In the region of middle stress with stress exponent n ∼ 3 and activation energy Q ∼ (195±7) kJ.mol -1 , micro-creep, caused by viscous gliding of dislocations due to the interaction of O atoms with dislocations, controls the deformation. In the high stress region with stress exponent n ∼ 5-6 and activation energy Q ∼ (210±10) kJ.mol -1 , two mechanisms of the climb of edge dislocations (EDC) and the motion of jogged screw dislocation (MJS) contribute to rate controlling process. In test conditions N36 alloy cladding tube behaves a type of creep similar to that noted in class-I (A) alloys

  20. Experimental study and modeling of high-temperature oxidation and phase transformation of cladding-tubes made in zirconium alloy

    International Nuclear Information System (INIS)

    Mazeres, Benoit

    2013-01-01

    One of the hypothetical accident studied in the field of the safety studies of Pressurized light Water Reactor (PWR) is the Loss-Of-Coolant-Accident (LOCA). In this scenario, zirconium alloy fuel claddings could undergo an important oxidation at high temperature (T≅ 1200 C) in a steam environment. Cladding tubes constitute the first confinement barrier of radioelements and then it is essential that they keep a certain level of ductility after quenching to ensure their integrity. These properties are directly related to the growth kinetics of both the oxide and the αZr(O) phase and also to the oxygen diffusion profile in the cladding tube after the transient. In this context, this work was dedicated to the understanding and the modeling of the both oxidation phenomenon and oxygen diffusion in zirconium based alloys at high temperature. The numerical tool (EKINOX-Zr) used in this thesis is based on a numerical resolution of a diffusion/reaction problem with equilibrium-conditions on three moving boundaries: gas/oxide, oxide/αZr(O), αZr(O)/βZr. EKINOX-Zr kinetics model is coupled with ThermoCalc software and the Zircobase database to take into account the influence of the alloying elements (Sn, Fe, Cr, Nb) but also the influence of hydrogen on the solubility of oxygen. This study focused on two parts of the LOCA scenario: the influence of a pre-oxide layer (formed in-service) and the effects of hydrogen. Thanks to the link between EKINOX-Zr and the thermodynamic database Zircobase, the hydrogen effects on oxygen solubility limit could be considered in the numerical simulations. Thus, simulations could reproduce the oxygen diffusion profiles measured in pre-hydrided samples. The existence of a thick pre-oxide layer on cladding tubes can induce a reduction of this pre-oxide layer before the growth of a high-temperature one during the high temperature dwell under steam. The first simulations performed using the numerical tool EKINOX-Zr showed that this particular

  1. 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

  2. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes: Flaw detection and characterization [Phase 1

    International Nuclear Information System (INIS)

    2006-05-01

    Nuclear power plants with heavy water reactors (HWRs) comprise nine percent of today's operating nuclear units, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems 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 Programme (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the frame of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of the IAEA's project on advanced technologies for HWRs. The objective of the CRP was to inter-compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP, participants have investigated the capability of different techniques to detect and characterize flaws. During the second phase of this CRP, participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in Zirconium alloys. The intent was to identify the most effective pressure tube inspection and diagnostic methods, and to identify further development needs. The organizations that have participated in this CRP are: - The Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL); Chalk River Laboratories (CRL), Canada; - The Research Institute of Nuclear Power Operations (RINPO), China National Nuclear Corporation (CNNC), China; - Bhabha Atomic Research Centre (BARC), India; - The Korea Electric Power Research Institute (KEPRI), Republic of Korea; - The Korea Atomic Energy

  3. Susceptibility of cold-worked zirconium-2.5 wt% niobium alloy to delayed hydrogen cracking

    International Nuclear Information System (INIS)

    Coleman, C.E.

    1976-01-01

    Notched tensile specimens of cold-worked zirconium-2.5 wt% niobium alloy have been stressed at 350 K and 520 K. At 350 K, above a possible threshold stress of 200 MPa, specimens exhibited delayed failure which was attributed to hydride cracking. Metallography showed that hydrides accumulated at notches and tips of growing cracks. The time to failure appeared to be independent of hydrogen content over the range 7 to 100 ppm hydrogen. Crack growth rates of about 10 -10 m/s deduced from fractography were in the same range as those necessary to fracture pressure tubes. The asymptotic stress intensity for delayed failure, Ksub(1H), appeared to be about 5 MPa√m. With this low value of Ksub(1H) small surface flaws may propagate in pressure tubes which contain large residual stresses. Stress relieving and modified rolling procedures will reduce the residual stresses to such an extent that only flaws 12% of the wall thickness or greater will grow. At 520 K no failures were observed at times a factor of three greater than times to failure at 350 K. Zirconium-2.5 wt% niobium appears to be safe from delayed hydrogen cracking at the reactor operating temperature. (author)

  4. 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)

  5. Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar [University of Wisconsin-Madison; Mariani, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Company; Lahoda, Ed [Westinghouse Electric Company

    2018-03-31

    Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi2 coating) during clad quenching experiments is discussed in detail. Oxidation of bulk ZrSi2 was found to be negligible compared to Zircaloy-4 (a common Zr-alloy cladding material) and mechanical integrity of ZrSi2 was superior to that of bulk Zr2Si at high temperatures in ambient air. Very interesting and unique multi-nanolayered composite of ZrO2 and SiO2 were observed. Physical model for the oxidation has been proposed wherein Zr–Si–O mixture undergoes a spinodal phase decomposition into ZrO2 and SiO2, which is manifested as a nanoscale assembly of alternating layer of the two oxides. Steam corrosion at high pressure (10.3 MPa) led to weight loss of ZrSi2 and produced oxide scale with depletion of silicon, possibly attributed to volatile silicon hydroxide, gaseous silicon monoxide, and a solubility of silicon dioxide in water. Only Zircon phase (ZrSiO4) formed during oxidation of ZrSi2 at 1400°C in air, and allowed for immobilization silicon species in oxide scale in the aqueous environments. Zirconium-silicide coatings (on zirconium-alloy substrates) investigated in this study were deposited primarily using magnetron sputter deposition method and slurry method, although powder spray deposition processes cold spray and thermal spray methods were also investigated. The optimized ZrSi2 sputtered coating exhibited a highly protective nature at elevated temperatures in ambient air by mitigating oxygen permeation to the underlying zirconium alloy substrate. The high oxidation resistance of the coating has been shown to be due to nanocrystalline SiO2 and ZrSiO4 phases in the amorphous

  6. 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)

  7. Process for purifying zirconium sponge

    International Nuclear Information System (INIS)

    Abodishish, H.A.M.; Kimball, L.S.

    1992-01-01

    This patent describes a Kroll reduction process wherein a zirconium sponge contaminated with unreacted magnesium and by-product magnesium chloride is produced as a regulus, a process for purifying the zirconium sponge. It comprises: distilling magnesium and magnesium chloride from: a regulus containing a zirconium sponge and magnesium and magnesium chloride at a temperature above about 800 degrees C and at an absolute pressure less than about 10 mmHg in a distillation vessel to purify the zirconium sponge; condensing the magnesium and the magnesium chloride distilled from the zirconium sponge in a condenser; and then backfilling the vessel containing the zirconium sponge and the condenser containing the magnesium and the magnesium chloride with a gas; recirculating the gas between the vessel and the condenser to cool the zirconium sponge from above about 800 degrees C to below about 300 degrees C; and cooling the recirculating gas in the condenser containing the condensed magnesium and the condensed magnesium chloride as the gas cools the zirconium sponge to below about 300 degrees C

  8. The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components Delayed Hydride Cracking

    CERN Document Server

    Puls, Manfred P

    2012-01-01

    By drawing together the current theoretical and experimental understanding of the phenomena of delayed hydride cracking (DHC) in zirconium alloys, The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components: Delayed Hydride Cracking provides a detailed explanation focusing on the properties of hydrogen and hydrides in these alloys. Whilst the focus lies on zirconium alloys, the combination of both the empirical and mechanistic approaches creates a solid understanding that can also be applied to other hydride forming metals.   This up-to-date reference focuses on documented research surrounding DHC, including current methodologies for design and assessment of the results of periodic in-service inspections of pressure tubes in nuclear reactors. Emphasis is placed on showing that our understanding of DHC is supported by progress across a broad range of fields. These include hysteresis associated with first-order phase transformations; phase relationships in coherent crystalline metallic...

  9. 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)

  10. 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.

  11. 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

  12. 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.

  13. 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

  14. 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.

  15. 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

  16. 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).

  17. 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)

  18. Change in the work function of zirconium by oxidation at high temperatures and low oxygen pressures

    International Nuclear Information System (INIS)

    Maeno, Yutaka; Yamamoto, Masahiro; Naito, Shizuo; Mabuchi, Mahito; Hashino, Tomoyasu

    1991-01-01

    Changes in the work function of zirconium on oxidation are measured at oxygen pressures of 3.0 x 10 -6 - 3.0 x 10 -4 Pa and at temperatures in the range 426-775 K. The work function first decreases then increases until a final saturation stage is reached. Use of secondary-ion mass spectroscopy (SIMS) shows that the changes correspond to oxygen adsorption, oxide nucleation and oxide growth, respectively. The initial decrease in work function is interpreted by the incorporation of oxygen adatoms into the subsurface. The oxygen adsorption potential of zirconium is evaluated by an effective medium theory, and the physical origin of the incorporation of oxygen adatoms is discussed. The positive change in the work function caused by oxide formation and the temperature and pressure dependences of the change in the work function by oxidation are explained qualitatively. (author)

  19. 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)

  20. 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.

  1. 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

  2. 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.

  3. 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

  4. Deformation mechanisms and irradiation effects in zirconium alloys. A multi-scale study

    International Nuclear Information System (INIS)

    Onimus, Fabien

    2015-01-01

    Zirconium alloys have been used for more than 30 years in the nuclear industry as structural materials for the fuel assemblies of pressurized water reactors. In particular, the cladding tube, made of zirconium alloys, constitutes the first barrier against the dissemination of radioactive elements. It is therefore essential to have a good understanding and prediction of the mechanical behavior of these materials in various conditions. The work presented in this dissertation deals with an experimental study and numerical simulations, at several length scales, of the deformation mechanisms and the mechanical behavior of zirconium alloys before irradiation, but also after irradiation and under irradiation. The mechanical behavior of zirconium single crystal has been determined, during an original study, using tensile test specimens containing large grains. Based on this study, crystal plasticity constitutive laws have been proposed. A polycrystalline model has also been developed to simulate the behavior of unirradiated zirconium alloys. A thorough Transmission Electron Microscopy (TEM) study has been able to clarify the deformation mechanisms of zirconium alloys occurring after irradiation. The clearing of loops by gliding dislocations leading to the dislocation channeling mechanism has been studied in details. This phenomenon has also been simulated using a dislocation dynamics code. The macroscopic consequences of this process have also been analyzed. A polycrystalline model taking into account the specificity of this mechanism has eventually been proposed. This approach has then been extended to the post-irradiation creep behavior. The recovery of radiation defects during creep tests has been characterized by TEM and modeled using cluster dynamics method. Deformation modes during creep have also been studied and a simple model for the creep behavior has eventually been proposed. Finally, the mechanism responsible for the acceleration of irradiation growth that

  5. 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)

  6. 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

  7. Tensile strength of Zr-2.5 Nb pressure tubes: A statistical study

    Energy Technology Data Exchange (ETDEWEB)

    Shah, Priti Kotak, E-mail: pritik@barc.gov.in [Senior Scientist, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Dubey, J.S.; Datta, D.; Shriwastaw, R.S.; Rath, B.N.; Singh, R.N. [Senior Scientist, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Anantharaman, S. [Head, Post Irradiation Examination Division, Bhabha Atomic Research Centre, Mumbai (India); Chakravartty, J.K. [Director, Materials Group, Bhabha Atomic Research Centre, Mumbai (India)

    2015-12-15

    Highlights: • Tensile properties in axial and transverse direction for a number of Indian Zr-2.5 Nb PHWR pressure tubes. • Distribution of tensile properties of double-melted and quadruple-melted pressure tubes. • Tensile properties at front-end and back-end of the quadruple melted pressure tubes at room temperature and at 300 °C. - Abstract: In order to get an idea of the statistical variation in the tensile properties of the double-melted as well as quadruple melted Zr-2.5 Nb pressure tubes (PTs) and also the variation in tensile properties between the two ends of the pressure tubes, tension tests were carried out on around 50 pressure tube off-cuts. Longitudinal and transverse tensile specimens were prepared from these off-cuts of pressure tubes of double-melted and quadruple melted types. For quadruple melted pressure tubes the specimens were tested from both front-end and back-end off-cuts. Miniature flat tensile specimens having 1.8 mm width and 1.5 mm thickness and 7.6 mm gauge length were prepared from the pressure tube off-cuts without any flattening treatment. Tension tests were carried out in a screw-driven machine at room temperature and 300 °C for both front-end and back-end off-cuts of each of 16 pressure tubes. In general the transverse specimens showed higher yield strength (YS) and ultimate tensile strength (UTS) compared to the longitudinal specimens. Transverse specimens showed less strain hardening compared to the longitudinal specimens. The axial specimens showed higher uniform (UE) and total elongation (TE) compared to the transverse specimens. Double-melted pressure tubes showed relatively higher strength and lower elongation and larger standard deviation compared to the quadruple melted pressure tubes. Mean values of tensile properties showed that back-end off-cuts were relatively stronger and less ductile compared to the front-end off-cuts.

  8. MAX Phase Modified SiC Composites for Ceramic-Metal Hybrid Cladding Tubes

    International Nuclear Information System (INIS)

    Jung, Yang-Il; Kim, Sun-Han; Park, Dong-Jun; Park, Jeong-Hwan; Park, Jeong-Yong; Kim, Hyun-Gil; Koo, Yang-Hyun

    2015-01-01

    A metal-ceramic hybrid cladding consists of an inner zirconium tube, and an outer SiC fiber-matrix SiC ceramic composite with surface coating as shown in Fig. 1 (left-hand side). The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. In addition, the outermost layer prevents the dissolution of SiC during normal operation. On the other hand, a ceramic-metal hybrid cladding consists of an outer zirconium tube, and an inner SiC ceramic composite as shown in Fig. 1 (right-hand side). The outer zirconium protects the fuel rod from a corrosion during reactor operation, as in the present fuel claddings. The inner SiC composite, additionally, is designed to resist the severe oxidation under a postulated accident condition of a high-temperature steam environment. Reaction-bonded SiC was fabricated by modifying the matrix as the MAX phase. The formation of Ti 3 SiC 2 was investigated depending on the compositions of the preform and melt. In most cases, TiSi 2 was the preferential phase because of its lowest melting point in the Ti-Si-C system. The evidence of Ti 3 SiC 2 was the connection with the pressurizing

  9. Spring/dimple instrument tube restraint

    International Nuclear Information System (INIS)

    DeMario, E.E.; Lawson, C.N.

    1993-01-01

    A nuclear fuel assembly for a pressurized water nuclear reactor has a spring and dimple structure formed in a non-radioactive insert tube placed in the top of a sensor receiving instrumentation tube thimble disposed in the fuel assembly and attached at a top nozzle, a bottom nozzle, and intermediate grids. The instrumentation tube thimble is open at the top, where the sensor or its connection extends through the cooling water for coupling to a sensor signal processor. The spring and dimple insert tube is mounted within the instrumentation tube thimble and extends downwardly adjacent the top. The springs and dimples restrain the sensor and its connections against lateral displacement causing impact with the instrumentation tube thimble due to the strong axial flow of cooling water. The instrumentation tube has a stainless steel outer sleeve and a zirconium alloy inner sleeve below the insert tube adjacent the top. The insert tube is relatively non-radioactivated inconel alloy. The opposed springs and dimples are formed on diametrically opposite inner walls of the insert tube, the springs being formed as spaced axial cuts in the insert tube, with a web of the insert tube between the cuts bowed radially inwardly for forming the spring, and the dimples being formed as radially inward protrusions opposed to the springs. 7 figures

  10. 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

  11. 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)

  12. 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.

  13. 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

  14. 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

  15. 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.

  16. 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

  17. 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)

  18. Creep behavior under internal pressure of zirconium alloy cladding oxidized in steam at high temperature

    International Nuclear Information System (INIS)

    Chosson, Raphael

    2014-01-01

    During hypothetical Loss-Of-Coolant-Accident (LOCA) scenarios, zirconium alloy fuel cladding tubes creep under internal pressure and are oxidized on their outer surface at high temperature (HT). Claddings become stratified materials: zirconia and oxygen-stabilized α phase, called α(O), are formed on the outer surface of the cladding whereas the inner part remains in the β domain. The strengthening effect of oxidation on the cladding creep behavior under internal pressure has been highlighted at HT. In order to model this effect, the creep behavior of each layer had to be determined. This study focused on the characterization of the creep behavior of the α(O) phase at HT, through axial creep tests performed under vacuum on model materials, containing from 2 to 7 wt.% of oxygen and representative of the α(O) phase. For the first time, two creep flow regimes have been observed in this phase. Underlying physical mechanisms and relevant microstructural parameters have been discussed for each regime. The strengthening effect due to oxygen on the α(O) phase creep behavior at HT has been quantified and creep flow equations have been identified. A ductile to brittle transition criterion has been also suggested as a function of temperature and oxygen content. Relevance of the creep flow equations for each layer, identified in this study or from the literature, has been discussed. Then, a finite element model, describing the oxidized cladding as a stratified material, has been built. Based on this model, a fraction of the experimental strengthening during creep is predicted. (author) [fr

  19. 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

  20. Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

    2010-01-01

    In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

  1. 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

  2. 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

  3. 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

  4. Experimental and thermodynamic study of the erbium-oxygen-zirconium and gadolinium-oxygen-zirconium systems

    International Nuclear Information System (INIS)

    Jourdan, J.

    2009-11-01

    This work is a contribution to the development of innovative concepts for fuel cladding in pressurized water nuclear reactors. This concept implies the insertion of rare earth (erbium and gadolinium) in the zirconium fuel cladding. The determination of phase equilibria in the systems is essential prior to the implementation of such a promising solution. This study consisted in an experimental determination of the erbium-zirconium phase diagram. For this, we used many different techniques in order to obtain diagram data such as solubility limits, solidus, liquidus or invariant temperatures. These data allowed us to present a new diagram, very different from the previous one available in the literature. We also assessed the diagram using the CALPHAD approach. In the gadolinium-zirconium system, we determined experimentally the solubility limits. Those limits had never been determined before, and the values we obtained showed a very good agreement with the experimental and assessed versions of the diagram. Because these alloys are subjected to oxygen diffusion throughout their life, we focused our attention on the erbium-oxygen-zirconium and gadolinium-oxygen-zirconium systems. The first system has been investigated experimentally. The alloys fabrication has been performed using powder metallurgy. In order to obtain pure raw materials, we fabricated powder from erbium and zirconium bulk metals using hydrogen absorption/desorption. The characterisation of the ternary pellets allowed the determination of two ternary isothermal sections at 800 and 1100 C. For the gadolinium-oxygen-zirconium system, we calculated the phase equilibria at temperatures ranging from 800 to 1100 C, using a homemade database compiled from literature assessments of the oxygen-zirconium, gadolinium-zirconium and gadolinia-zirconia systems. Finally, we determined the mechanical properties, in connexion with the microstructure, of industrial quality alloys in order to identify the influence of

  5. 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)

  6. 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.

  7. 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

  8. The chemical vapor deposition of zirconium carbide onto ceramic substrates

    International Nuclear Information System (INIS)

    Glass A, John Jr.; Palmisiano, Nick Jr.; Welsh R, Edward

    1999-01-01

    Zirconium carbide is an attractive ceramic material due to its unique properties such as high melting point, good thermal conductivity, and chemical resistance. The controlled preparation of zirconium carbide films of superstoichiometric, stoichiometric, and substoichiometric compositions has been achieved utilizing zirconium tetrachloride and methane precursor gases in an atmospheric pressure high temperature chemical vapor deposition system

  9. 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

  10. Anisotropic deformation of Zr–2.5Nb pressure tube material at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Fong, R.W.L., E-mail: fongr@aecl.ca [Fuel and Fuel Channel Safety Branch, Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories, Chalk River, Ontario (Canada)

    2013-09-15

    Zr–2.5Nb alloy is used for the pressure tubes in CANDU® reactor fuel channels. In reactor, the pressure tube normally operates at 300 °C and experiences a primary coolant fluid internal pressure of approximately 10 MPa. Manufacturing and processing procedures generate an anisotropic state in the pressure tube which makes the tube stronger in the hoop (transverse) direction than in the axial (longitudinal) direction. This anisotropy condition is present for temperatures less than 500 °C. During postulated accident conditions where the material temperature could reach 1000 °C, it might be assumed that the high temperature and subsequent phase change would reduce the inherent anisotropy, and thus affect the deformation behaviour (ballooning) of the pressure tube. From constant-load, rapid-temperature-ramp, uniaxial deformation tests, the deformation rate in the longitudinal direction of the tube behaves differently than the deformation rate in the transverse direction of the tube. This anisotropic mechanical behaviour appears to persist at temperatures up to 1000 °C. This paper presents the results of high-temperature deformation tests using longitudinal and transverse specimens taken from as-received Zr–2.5Nb pressure tubes. It is shown that the anisotropic deformation behaviour observed at high temperatures is largely due to the stable crystallographic texture of the α-Zr phase constituent in the material that was previously observed by neutron diffraction measurements during heating at temperatures up to 1050 °C. The deformation behaviour is also influenced by the phase transformation occurring at high temperatures during heating. The effects of texture and phase transformation on the anisotropic deformation of as-received Zr–2.5Nb pressure tube material are discussed in the context of the tube ballooning behaviour. Because of the high temperatures in postulated accident scenarios, any irradiation damage will be annealed from the pressure tube material

  11. Mitigation of harmful effects of welds in zirconium alloy components

    International Nuclear Information System (INIS)

    Coleman, C.E.; Doubt, G.L.; Fong, R.W.L.; Root, J.H.; Bowden, J.W.; Sagat, S.; Webster, R.T.

    1995-01-01

    Welding produces local residual tensile stresses and changes in texture in components made from zirconium alloys. In the heat-affected zone in tubes or plates, the basal plane normals are rotated into the plane of the component and perpendicular to the direction of the weld. Thin-walled Zircaloy-2 tubes containing an axial weld do not reach their full strength because they always fail prematurely in the weld when pressurized to failure in a fixed-end burst test. Reinforcing the weld by increasing its thickness by 25% moves the failure to the parent metal and improves the biaxial strength of the tube by 20 to 25% and increases the total elongation by 200 to 450%. In components made from Zr-2.5Nb, the texture in the heat-affected zone promotes delayed hydride cracking (DHC) driven by tensile residual stress. Although the texture is not much affected by heat-treatments below 630 o C and large grain interaction stresses remain as a result of mixed textures, macro-residual tensile stresses can be relieved by heat treatment to the point where the probability of cracking is very low. (author)

  12. Mitigation of harmful effects of welds in zirconium alloy components

    International Nuclear Information System (INIS)

    Coleman, C.E.; Doubt, G.L.; Fong, R.W.L.; Root, J.H.; Bowden, J.W.; Sagat, S.

    1994-01-01

    Welding produces local residual tensile stresses and changes in texture in components made from zirconium alloys. In the heat-affected zone in tubes or plates, the basal plane normals are rotated into the plane of the component and perpendicular to the direction of the weld. Thin-walled Zircaloy-2 tubes containing an axial weld do not reach their full strength because they always fail prematurely in the weld when pressurized to failure in a fixed-end burst test. Reinforcing the weld by increasing its thickness by 25% moves the failure to the parent metal and improves the biaxial strength of the tube by 20 to 25% and increases the total elongation by 200 to 450%. In components made from Zr-2.5Nb, the texture in the heat-affected zone promotes delayed hydride cracking (DHC) driven by tensile residual stress. Although the texture is not much affected by heat-treatments below 630 C and large grain interaction stresses remain as a result of mixed textures, macro-residual tensile stresses can be relieved by heat treatment to the point where the probability of cracking is very low

  13. 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)

  14. 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.)

  15. 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

  16. 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.)

  17. 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

  18. 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....

  19. 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....

  20. 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

  1. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    International Nuclear Information System (INIS)

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong

    2012-01-01

    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced

  2. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced.

  3. 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.

  4. 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

  5. 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

  6. 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

  7. Highly corrosion resistant zirconium based alloy for reactor structural material

    International Nuclear Information System (INIS)

    Ito, Yoichi.

    1996-01-01

    The alloy of the present invention is a zirconium based alloy comprising tin (Sn), chromium (Cr), nickel (Ni) and iron (Fe) in zirconium (Zr). The amount of silicon (Si) as an impurity is not more than 60ppm. It is preferred that Sn is from 0.9 to 1.5wt%, that of Cr is from 0.05 to 0.15wt%, and (Fe + Ni) is from 0.17 to 0.5wt%. If not less than 0.12wt% of Fe is added, resistance against nodular corrosion is improved. The upper limit of Fe is preferably 0.40wt% from a view point of uniform suppression for the corrosion. The nodular corrosion can be suppressed by reducing the amount of Si-rich deposition product in the zirconium based alloy. Accordingly, a highly corrosion resistant zirconium based alloy improved for the corrosion resistance of zircaloy-2 and usable for a fuel cladding tube of a BWR type reactor can be obtained. (I.N.)

  8. 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.)

  9. 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)

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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)

  16. Development of zirconium hydride highly effective moderator materials

    International Nuclear Information System (INIS)

    Yin Changgeng

    2005-10-01

    The zirconium hydride with highly content of hydrogen and low density is new efficient moderator material for space nuclear power reactor. Russia has researched it to use as new highly moderator and radiation protection materials. Japanese has located it between the top of pressure vessel and the main protection as a shelter, the work temperature is rach to 220 degree C. The zirconium hydride moderator blocks are main parts of space nuclear power reactor. Development of zirconium hydride moderator materials have strength research and apply value. Nuclear Power Research and Design Instituteoh China (NPIC) has sep up the hydrogenation device and inspect systems, and accumurate a large of experience about zirconium hydride, also set up a strict system of QA and QC. (authors)

  17. Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys; Approche experimentale et modelisation micromecanique du comportement des alliages de zirconium irradies

    Energy Technology Data Exchange (ETDEWEB)

    Onimus, F

    2003-12-01

    Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

  18. Delayed hydride cracking in irradiated Zr-2.5 % Nb pressure tubes

    International Nuclear Information System (INIS)

    Cirimello, Pablo; Coronel, Pascual; Haddad, Roberto; Lafont, Claudio; Mizrahi, Rafael

    2003-01-01

    Pressure tubes in CANDU nuclear power plants are made of Zr-2.5 % Nb alloy, which is susceptible to a cracking process called Delayed Hydride Cracking (DHC). Measurement of DHC velocity on irradiated pressure tubes is essential to assure the validity of the Leak Before Break criterion. This work was performed on samples from two pressure tubes taken out of the Embalse NPP in 1995, belonging to fuel channels A-14 and L-12. DHC velocity in the axial direction was measured at 211 C degrees for samples taken from different axial positions, which allowed to study its dependence on fast neutron fluency and irradiation temperature. Non-irradiated material was also tested. It was found that DHC velocity results for the tested material were similar to those obtained for a great number of tubes irradiated in other CANDU plants. (author)

  19. Correlation between the critical heat flux and the fractal surface roughness of zirconium alloy tubes

    International Nuclear Information System (INIS)

    Fong, R.W.L.; McRae, G.A.; Coleman, C.E.; Nitheanandan, T.; Sanderson, D.B.

    1999-10-01

    In CANDU fuel channels, Zircaloy calandria tubes isolate the hot pressure tubes from the cool heavy water moderator. The heavy-water moderator provides a backup heat sink during some postulated loss-of-coolant accidents. The decay heat from the fuel is transferred to the moderator to ensure fuel channel integrity during emergencies. Moderator temperature requirements are specified to ensure that the transfer of decay heat does not exceed the critical heat flux (CHF) on the outside surface of the calandria tube. An enhanced CHF provides increases in safety margin. Pool boiling experiments indicate the CHF is enhanced with glass-peening of the outside surface of the calandria tubes. The objective of this study was to evaluate the surface characteristics of glass-peened tubes and relate these characteristics to CHF. The micro-topologies of the tube surfaces were analysed using stereo-pair micrographs obtained from scanning electron microscopy (SEM) and photogrammetry techniques. A linear relationship correlated the CHF as a function of the 'fractal' surface roughness of the tubes. (author)

  20. 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

  1. 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

  2. [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.

  3. Electroless deposition process for zirconium and zirconium alloys

    Science.gov (United States)

    Donaghy, Robert E.; Sherman, Anna H.

    1981-01-01

    A method is disclosed for preventing stress corrosion cracking or metal embrittlement of a zirconium or zirconium alloy container that is to be coated on the inside surface with a layer of a metal such as copper, a copper alloy, nickel, or iron and used for holding nuclear fuel material as a nuclear fuel element. The zirconium material is etched in an etchant solution, desmutted mechanically or ultrasonically, oxidized to form an oxide coating on the zirconium, cleaned in an aqueous alkaline cleaning solution, activated for electroless deposition of a metal layer and contacted with an electroless metal plating solution. This method provides a boundary layer of zirconium oxide between the zirconium container and the metal layer.

  4. Method to electrolytically precipitate metals onto zirconium objects

    International Nuclear Information System (INIS)

    Donaghy, R.E.

    1978-01-01

    Tubes and other formed bodies made of zirconium or zirconium alloys which serve to take up nuclear fuels, are plated by electrolytically depositing a metal film onto these in order to improve their mechanical and corrosion properties. The object is activated in a solution of ammonium bifluoride and sulphuric acid, whereby an electrically conducting solid and a loose layer is formed. This loose film is removed by using fluoboric acid or hydrofluoric silicic acid solution, ultrasonics, or strips of organic material (cotton, polyester, nylon). The plating of Cu, Ni, Cr is described in detail. The object is rinsed between the process steps with deionized water and finally degased at a temperature of 150-200 0 C. (IHOE) [de

  5. Molten salt extractive distillation process for zirconium-hafnium separation

    International Nuclear Information System (INIS)

    McLaughlin, D.F.; Stoltz, R.A.

    1989-01-01

    This patent describes an improvement in a process for zirconium-hafnium separation. It utilizes an extractive distillation column with a mixture of zirconium and hafnium tetrachlorides introduced into a distillation column having a top and bottom with hafnium enriched overheads taken from the top of the column and a molten salt solvent circulated through the column to provide a liquid phase, and with molten salt solvent containing zirconium chloride being taken from the bottom of the distillation column. The improvements comprising: utilizing a molten salt solvent consisting principally of lithium chloride and at least one of sodium, potassium, magnesium and calcium chlorides; stripping of the zirconium chloride taken from the bottom of the distillation column by electrochemically reducing zirconium from the molten salt solvent; and utilizing a pressurized reflux condenser on the top of the column to add the hafnium chloride enriched overheads to the molten salt solvent previously stripped of zirconium chloride

  6. 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.

  7. Identification of the zirconium hydrides metallography in zircaloy-2

    International Nuclear Information System (INIS)

    Garcia Gonzalez, F.

    1968-01-01

    Technique for the Identification of the zirconium hydrides in metallographic specimens have been developed. Microhardness, quantitative estimation and relative orientation of the present hydrides as well as grain size determination of the different Zircaloy-2 tube specimens have also been made. The specimens used were corrosion- tested in water during various periods of time at 300 degree castrating, prior to the metallographic examination. Reference specimens, as received, and heavily hydride specimens in a hydrogen atmosphere at 800 degree centigrees, have been used in the previous stages of the work. No difficulties have been met in this early stage of acquaintanceship with the zirconium hydrides. (Author) 5 refs

  8. 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.)

  9. Lithium aluminate/zirconium material useful in the production of tritium

    Science.gov (United States)

    Cawley, W.E.; Trapp, T.J.

    A composition is described useful in the production of tritium in a nuclear reactor. Lithium aluminate particles are dispersed in a matrix of zirconium. Tritium produced by the reactor of neutrons with the lithium are absorbed by the zirconium, thereby decreasing gas pressure within capsules carrying the material.

  10. 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

  11. 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.

  12. 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

  13. Delayed hydride cracking in zirconium alloys in pressure tube nuclear reactors. Final report of a coordinated research project 1998-2002

    International Nuclear Information System (INIS)

    2004-10-01

    This report describes all of the research work undertaken as part of the IAEA coordinated research project on hydrogen and hydride induced degradation of the mechanical and physical properties of zirconium based alloys, and includes a review of the state of the art in understanding crack propagation by Delayed Hydride Cracking (DHC), and details of the experimental procedures that have produced the most consistent set of DHC rates reported in an international round-robin exercise to this date. It was concluded that 1) the techniques for performing measurements of the rate of delayed hydride cracking in zirconium alloys have been transferred from the host laboratory to other countries; 2) by following a strict procedure, a very consistent set of values of crack velocity were obtained by both individual laboratories and between the different laboratories; 3) the results over a wide range of test temperatures from materials with various microstructures fitted into the current theoretical framework for delayed hydride cracking; 4) an inter-laboratory comparison of hydrogen analysis revealed the importance of calibration and led to improvements in measurement in the participating laboratories and 5) the success of the CRP in achieving its goals has led to the initiation of some national programmes

  14. 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....

  15. 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)

  16. Traps in Zirconium Alloys Oxide Layers

    Directory of Open Access Journals (Sweden)

    Helmar Frank

    2005-01-01

    Full Text Available Oxide films long-time grown on tubes of three types of zirconium alloys in water and in steam were investigated, by analysing I-V characteristic measured at constant voltages with various temperatures. Using theoretical concepts of Rose [3] and Gould [5], ZryNbSn(Fe proved to have an exponential distribution of trapping centers below the conduction band edge, wheras Zr1Nb and IMP Zry-4 proved to have single energy trap levels.

  17. 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

  18. 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

  19. Investigation on the corrosion resistance of zirconium in nitric acid

    International Nuclear Information System (INIS)

    Fauvet, P.; Mur, P.

    1990-01-01

    Zirconium in nitric solutions exhibits an excellent corrosion resistance in the passive state, and a mediocre corrosion resistance in the unpassive state with risk of stress corrosion cracking. Results of the influence of some parameters (medium, potential, temperature, stress, friction, metallurgical structure and surface state) on zirconium passivation are presented. Zirconium remains passive in a large range of HNO 3 concentration (at least up to 14.4N), in the presence of oxidizing ions (Cr 4 , Ce 4 ), in a spent fuel dissolution solution. Zirconium is depassived by friction at high speed and pressure, by platinum coupling in boiling 14.4N HNO 3 with or without stress, or by imposed deformation speed under high potential. (A.B.)

  20. Device and method for shortening reactor process tubes

    Science.gov (United States)

    Frantz, Charles E.; Alexander, William K.; Lander, Walter E. B.

    1980-01-01

    This disclosure describes a device and method for in situ shortening of nuclear reactor zirconium alloy process tubes which have grown as a result of radiation exposure. An upsetting technique is utilized which involves inductively heating a short band of a process tube with simultaneous application of an axial load sufficient to cause upsetting with an attendant decrease in length of the process tube.

  1. 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.

  2. 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)

  3. Heat treated tube for cladding nuclear fuel element

    International Nuclear Information System (INIS)

    Eddens, F.C.; White, D.W.; Harmon, J.L.

    1983-01-01

    The zirconium alloy tube comprises a metallurgical gradient across the width of the tube wall wherein the tube has a more corrosion-resistant metallurgical condition at the outer circumference and a less corrosion-resistant metallurgical condition at the inner circumference. The metallurgical gradient can be generated by heating an outer circumferential portion of the tube to the high alpha or mixed alpha plus beta range while maintaining the inner surface at a lower temperature, followed by cooling of the tube. Preferably the tube is made of Zircaloy. (author)

  4. Fracture toughness of irradiated Zr-2.5Nb pressure tube from Indian PHWR

    Science.gov (United States)

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

    2015-03-01

    Fracture toughness of irradiated Zr-2.5Nb alloy pressure tube, fabricated by the cold pilgering and stress relieving route, was evaluated using disk compact tension type specimens. These specimens were punched out from the irradiated pressure tube (S-07), which was in service for about 8 effective full power years of reactor operation in the Kakrapar Atomic Power Station-2 (KAPS-2). The tests were carried out remotely inside a lead shielded enclosure. Crack growth during the test was measured using the direct current potential drop technique. The irradiated pressure tube showed low fracture toughness at 25 °C. The fracture toughness increased with increase in temperature up to 250 °C but was practically unaffected with further increase in temperature up to 300 °C. This paper discusses the fracture behavior of irradiated Indian pressure tube material and compares it with other data available.

  5. 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)

  6. 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

  7. Experimental study of the zirconium alloy oxidation under high pressure of steam and modelling of the mechanisms

    International Nuclear Information System (INIS)

    Dali, Yacoub

    2007-01-01

    The corrosion of the cladding materials used for the fuel rods is one of the limiting factor of their lifetime in light water reactors. In this field, the aim of the nuclear industry is today to increase the time and the number of cycles and to submit the claddings in zirconium alloys to higher corrosive conditions. In this way, new alloys devoted to replace the standard Zircaloy-4, for instance Nb containing alloys, have been recently developed and licensed and show better corrosion resistance. A better understanding of the corrosion mechanisms of the zirconium alloys is necessary to predict the corrosion behaviour of these materials. In this work, the oxidation rate of model alloys of two metallurgic families has been studied in steam in a pressure range between 100 milli-bars and 100 bars. The Zircaloy type alloys contain as alloying elements oxygen and/or tin and/or iron and chromium. For the Zr-Nb family, three niobium contents have been studied, respectively 0.2, 0.4 and 1 weight percent of niobium. Our objectives were to understand the variations of the reactivity between the low pressure and the high pressure range, in quantifying the dependency of the corrosion rate with the steam pressure and the alloying element concentrations. The segregation process of the niobium at the surface has also been studied on the Zr-Nb alloys. During this work, a magnetic suspension thermo-balance has been developed and used to follow in-situ the corrosion rate at high pressure of water vapour. The oxide layers have been characterized by many techniques, macro and micro-photo-electrochemistry, XRD, FEG-SEM, XPS, HR-TEM and SIMS. For the Zircaloy type alloys, we have confirmed the major role of the intermetallic precipitates Zr(Fe,Cr) 2 on the corrosion resistance. Unlike the standard Zircaloy-4, for which the oxidation rate does not depend on the pressure of the water vapour and is thus limited by the vacancy diffusion in the oxide layer, we have shown that the rate of the

  8. 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

  9. 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

  10. 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

  11. 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)

  12. Structural analysis of fuel rod applied to pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Danilo P.; Pinheiro, Andre Ricardo M.; Lotto, André A., E-mail: danilo.pinheiro@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil)

    2017-07-01

    The design of fuel assemblies applied to Pressurized Water Reactors (PWR) has several requirements and acceptance criteria that must be attended for licensing. In the case of PWR fuel rods, an important mechanical structural requirement is to keep the radial stability when submitted to the coolant external pressure. In the framework of the Accident Tolerant Fuel (ATF) program new materials have been studied to replace zirconium based alloys as cladding, including iron-based alloys. In this sense, efforts have been made to evaluate the behavior of these materials under PWR conditions. The present work aims to evaluate the collapse cold pressure of a stainless steel thin-walled tube similar to that used as cladding material of fuel rods by means of the comparison of numeric data, and experimental results. As a result of the simulations, it was observed that the collapse pressure has a value intermediate value between those found by regulatory requirements and analytical calculations. The experiment was carried out for the validation of the computational model using test specimens of thin-walled tubes considering empty tube. The test specimens were sealed at both ends by means of welding. They were subjected to a high pressure device until the collapse of the tubes. Preliminary results obtained from experiments with the empty test specimens indicate that the computational model can be validated for stainless steel cladding, considering the difference between collapse pressure indicated in the regulatory document and the actual limit pressure concerning to radial instability of tubes with the studied characteristics. (author)

  13. 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.

  14. 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)

  15. Tensile and fracture toughness characteristics of Zr-2.5Nb pressure tube

    International Nuclear Information System (INIS)

    Jung, H. C.; Kim, Y. S.; Ahn, S. B.; Kim, S. S.; Im, K. S.

    2004-01-01

    The object of this study is to evaluate the characteristics of tensile and fracture toughness of Zr-2.5Nb pressure tube. The transverse tensile tests were performed at various temperatures and the fracture toughness tests were carried out at room temperature using the CCT (curved compact tension) specimen. These specimens were directly machined from the pressure tube retaining original curvatures. Also, the fracture toughness of two sets of Zr-2.5Nb manufactured at different time was compared. The chemical analysis and the Vicker's hardness tests were performed at two sets of Zr-2.5Nb pressure tube. The Vicker's hardness value of SET-2 containing more oxygen and carbon relatively was higher about 11 than that of SET-1

  16. In-reactor creep of zirconium-2.5 wt% niobium at 570 K

    International Nuclear Information System (INIS)

    Coleman, C.E.; Causey, A.R.; Fidleris, V.

    1976-01-01

    The effect of fast neutron flux at 570 K on the creep rate of specimens of zirconium-2.5 wt% niobium alloy taken from tubes in various metallurgical conditions has been measured using both constant load tensile creep machines and bent-beam stress relaxation. Creep rates calculated from stress relaxation fit on the trend line for the constant load creep data. Between 114 MPa and 450 MPa the creep rate is proportional to neutron flux. The creep rate of specimens from the longitudinal direction is about twice that of specimens from the circumferential direction of a tube. This anisotropy in creep strength is attributed partly to crystallographic texture and partly to deformation substructure. Cold-work is detrimental to in-reactor creep strength; as-extruded material has higher creep strength. In cold-worked material at stresses below 100 MPa the stress exponent, n, is about 1; n gradually increases with stress being about 10 at 525 MPa and about 100 at 660 MPa. In laboratory tests, rupture ductility correlates inversely with n; the lower n the higher the ductility. In-reactor tests support this correlation thus pressure tubes in CANDU reactors, operating at 117 MPa where n approximately 1, should have good ductility. (Auth.)

  17. 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).

  18. Electrochemical stripping determination of traces of copper, lead, cadmium and zinc in zirconium metal and zirconium dioxide

    International Nuclear Information System (INIS)

    Stulik, K.; Beran, P.; Dolezal, J.; Opekar, F.

    1978-01-01

    Procedures have been developed for the determination of copper, lead, cadmium and zinc in zirconium metal and zirconium dioxide, at concentrations of 1ppm or less. Zirconium metal was dissolved in sulphuric acid, and zirconium dioxide decomposed under pressure with hydrofluoric acid. Sample solutions were prepared in dilute sulphuric acid. For the stripping determination, the sample solution was either mixed with a complexing tartrate base electrolyte or the pre-electrolysis was carried out in acid solution, with the acid solution being exchanged for a pure base electrolyte (e.g. an acetate buffer) for the stripping step. The stripping step was monitored by d.c., differential pulse and Kalousek commutator voltammetry and the three methods were compared. A stationary mercury-drop electrode can generally be used for all the methods, whereas a mercury-film electrode is suitable only for the d.c. voltammetric determination of copper, lead and cadmium, as pulse measurements with films are poorly reproducible and the electrodes are easily damaged. The relative standard deviation does not exceed 20%. Some samples contained relatively large amounts of copper, which is best separated by electrodeposition on a platinum electrode. (author)

  19. 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)

  20. 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.)

  1. 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.

  2. 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)

  3. 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.

  4. 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.

  5. 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

  6. 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.

  7. 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

  8. Synthesis of zirconium guanidinate complexes and the formation of zirconium carbonitride via low pressure CVD

    NARCIS (Netherlands)

    Potts, S.E.; Carmalt, C.J.; Blackman, C.S.; Abou-Chabine, F.; Pugh, D.; Davies, H.O.

    2009-01-01

    Thin films of zirconium carbonitride have been deposited on glass at 600 °C from two novel guanidinate precursors: [ZrCp'{¿2-(iPrN)2CNMe2}2Cl] (1) and [ZrCp'2{¿2-(iPrN)2CNMe2}Cl] (2) (Cp' ) monomethylcyclopentadienyl). Both compounds 1 and 2 were structurally characterized by X-ray crystallography.

  9. 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)

  10. 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.

  11. A layman's guide to radiation-induced deformation processes in zirconium alloys

    International Nuclear Information System (INIS)

    Dutton, R.

    1990-07-01

    The fuel channel (comprising a pressure tube and a calandria tube fabricated from zirconium alloys) in a CANDU reactor undergoes shape changes because of radiation-induced deformation. This is a consequence of the microstructural modification arising from radiation damage produced by the fast-neutron flux. This report summarizes our current understanding of the physical processes responsible for the deformation. With the non-specialist reader in mind, the underlying mechanisms are described in a manner that avoids much of the associated technical terminology. Thus, the basic concepts of plasticity in a crystalline material are introduced and related to the various microstructural defects created during irradiation. In particular, the mechanisms of creep (a time-dependent strain activated by an applied stress) and growth (a time-dependent strain occurring in the absence of stress) are discussed in a non-technical language assisted by simple diagrams. Reference is made to both theoretical investigations (avoiding mathematical complexity) and experimental measurements. It is shown how the qualitative and quantitative knowledge can be used to derive a predictive model for reactor designers and operators. The current status of such a model is evaluated and suggestions for future improvements made

  12. Hydrogen desorption kinetics from zirconium hydride and zirconium metal in vacuum

    International Nuclear Information System (INIS)

    Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.

    2014-01-01

    The kinetics of hydrogen desorption from zirconium hydride is important in many nuclear design and safety applications. In this paper, a coordinated experimental and modeling study has been used to explicitly demonstrate the applicability of existing kinetic theories for hydrogen desorption from zirconium hydride and α-zirconium. A static synthesis method was used to produce δ-zirconium hydride, and the crystallographic phases of the zirconium hydride were confirmed by X-ray diffraction (XRD). Three obvious stages, involving δ-zirconium hydride, a two-phase region, and α-zirconium, were observed in the hydrogen desorption spectra of two zirconium hydride specimens with H/Zr ratios of 1.62 and 1.64, respectively, which were obtained using thermal desorption spectroscopy (TDS). A continuous, one-dimensional, two-phase moving boundary model, coupled with the zero- and second-order kinetics of hydrogen desorption from δ-zirconium hydride and α-zirconium, respectively, has been developed to reproduce the TDS experimental results. A comparison of the modeling predictions with the experimental results indicates that a zero-order kinetic model is valid for description of hydrogen flux away from the δ-hydride phase, and that a second-order kinetic model works well for hydrogen desorption from α-Zr if the activation energy of desorption is optimized to be 70% of the value reported in the literature

  13. 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.)

  14. 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)

  15. Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys

    International Nuclear Information System (INIS)

    Onimus, F.

    2003-12-01

    Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

  16. 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.

  17. 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

  18. 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)

  19. Contribution to the understanding of zirconium alloy deformation under irradiation at high doses

    International Nuclear Information System (INIS)

    Gharbi, Nesrine

    2015-01-01

    The growth of zirconium alloy tubes of PWR fuel assemblies is the result of two phenomena: axial irradiation creep and stress 'free' growth which is correlated to the formation of c-loops at high irradiation doses. This PhD work aims at investigating the coupling between these two phenomena through a fine Transmission Electron Microscopy analysis of the effect of a macroscopic applied stress on the c-loop microstructure. 600 keV Zr + ion irradiations were performed at 300 C on two recrystallized zirconium alloys: Zircaloy-4 and M5. Thanks to a device specifically designed, different tensile or compressive stress levels were applied under ion irradiation. The microstructural observations have shown that the c-loop density reduces in grains oriented with the c-axis close to the direction of the applied tensile stress or far from the direction of the applied compressive stress, which is in good agreement with the SIPA mechanism. Nevertheless, the examination of a large number of grains has revealed dispersion from grain to grain. This dispersion, which could be explained by the intergranular heterogeneities, reduces the magnitude of the stress effect on c-loop microstructure. In parallel to this experimental study, a cluster dynamics model has been able to describe the evolution under irradiation of zirconium and Zircaloy-4 microstructure and to assess the effect of stress on c-loop microstructure. On the macroscopic scale, a physical model was also developed to predict the irradiation growth and creep behaviour of zirconium alloy tubes. (author) [fr

  20. 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...

  1. 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.

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. Axial and transverse tensile properties Zr-2.5Nb pressure tube off-cuts

    International Nuclear Information System (INIS)

    Shah, Priti K.; Dubey, J.S.; Shriwastaw, R.S.; Balakrishnan, K.S.; Anantharaman, S.; Chakravartty, J.K.

    2011-01-01

    Zr-2.5Nb alloys in cold worked and stress relieved (CWSR) condition serves as pressure boundary for hot coolant in Indian Pressurized Heavy Water Reactor (IPHWR). Due to both microstructural and crystallographic anisotropy, the mechanical properties in general and fracture behavior in particular are anisotropic for this material. To understand the anisotropic mechanical behavior of the Zr-2.5Nb pressure tubes of IPHWRs, tension tests were carried out on the RAPS-2 and KAPS-2 pressure tube off-cuts. Off-cuts are the small pieces cut from the two ends of a pressure tube before installing it into the PHWR. Miniature flat tensile specimens (without applying any flattening treatment to the pressure tube) of both longitudinal (or axial) and transverse orientation were fabricated from the off-cuts. Tension tests were carried out both at room temperature and 300 deg C. The transverse specimens showed higher strength and lower elongation than that of the axial ones. The back-end off-cuts showed higher strength compared to front-end off-cuts along axial direction whereas the same is not true for transverse direction specimens. One typical plot obtained in the testing of one of the off-cut is shown. The paper will discuss about the importance of the work carried out, test specimens, test method and results obtained in detail

  7. 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

  8. Manufacture of titanium and zirconium hydrides

    International Nuclear Information System (INIS)

    Mares, F.; Hanslik, T.

    1973-01-01

    A method is described of manufacturing titanium and zirconium hydrides by hydrogenation of said metals characterized by the reaction temperature ranging between 250 to 500 degC, hydrogen pressure of 20 to 300 atm and possibly by the presence of a hydride of the respective metal. (V.V.)

  9. INR participation in the IAEA research project investigating the influence of hydrogen absorption on zirconium alloy behavior

    International Nuclear Information System (INIS)

    Roth, Maria; Radu, Vasile; Dobrea, Dumitru; Pitigoi, Vasile

    2003-01-01

    The paper summarizes the results obtained at INR Pitesti from its participation in the research project coordinated by IAEA Vienna in cooperation with Chalk River and AECL Canada, titled 'Hydrogen and Hydride Induced Degradation of the Mechanical and Physical Properties of Zirconium-based Alloys'. Evidenced is the contribution of INR Pitesti in the works of this project as well as the benefits of this participation for Romania as owner of CANDU type reactor. In the frame this project new results concerning the propagation rate of DHC type cracks in pressure tubes in CANDU reactors were obtained. The same method used to investigate the DHC project was adapted for determination of other quantities of interest related to structural integrity of the materials. The methodology was applied for testing the pressure tubes in Cernavoda NPP Unit 1. The contribution of INR team to statistical processing of data obtained in all the laboratories participating in this project is also highlighted. Opportunity afforded by IAEA to INR Pitesti to bring its contribution to the development of this project of international cooperation together with other well-known institutions and the support from RAAN are acknowledged. These opened ways for other fruitful international cooperation

  10. 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)

  11. 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

  12. 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)

  13. Oxidation and deuterium uptake of Zr-2.5Nb pressure tubes in CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Urbanic, V.F.; Warr, B.D.; Manolescu, A.; Chow, C.K.; Shanahan, M.W.

    1989-01-01

    Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian deuterium uranium pressurized heavy-water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm 2 hydrogen per year) . The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes. (author)

  14. 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

  15. 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%.

  16. 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.)

  17. Precipitation of γ-zirconium hydride in zirconium

    International Nuclear Information System (INIS)

    Carpenter, G.J.C.

    1978-01-01

    A mechanism for the precipitation of γ-zirconium hydride in zirconium is presented which does not require the diffusion of zirconium. The transformation is completed by shears caused by 1/3 (10 anti 10) Shockley partial dislocations on alternate zirconium basal planes, either by homogeneous nucleation or at lattice imperfections. Homogeneous nucleation is considered least likely in view of the large nucleation barrier involved. Hydrides may form at dislocations by the generation of partials by means of either a pole or ratchet mechanism. The former requires dislocations with a component of Burgers vector along the c-axis, but contrast experiments show that these are not normally observed in annealed zirconium. It is therefore most likely that intragranular hydrides form at the regular 1/3 (11 anti 20) dislocations, possibly by means of a ratchet mechanism. Contrast experiments in the electron microscope show that the precipitates have a shear character consistent with the mechanism suggested. The possibility that the shear dislocations associated with the hydrides are emissary dislocations is considered and a model suggested in which this function is satisfied together with the partial relief of misfit stresses. The large shear strains associated with the precipitation mechanism may play an important role in the preferential orientation of hydrides under stress

  18. 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.

  19. 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

  20. 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.

  1. 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

  2. Research into zirconium alloys resistant to carbon dioxide under pressure at temperatures of up to 600 deg C (1963)

    International Nuclear Information System (INIS)

    Baque, P.; Dominget, R.; Bossard, J.

    1963-01-01

    Zirconium is a metal having a relatively low neutron capture cross-section and a high melting point; it is thus possible to consider its use in particular as a canning material for fuel elements in CO 2 -cooled nuclear reactors. A preliminary study of several types of zirconium showed that the metal is already strongly oxidised in this gas at 500 deg C. The 'breakaway' phenomenon is generalised; the oxidation rate is then linear and depends on the carbon dioxide pressure. An attempt was therefore made to find binary and tertiary alloys in order to improve the metal behaviour. Several interesting compositions were found: 1, 1.6 and 2.5 per cent of copper, 2 per cent of vanadium, and 0.05 and 0.5 per cent of calcium. Tertiary copper-molybdenum and copper-phosphorus alloys are also less liable to oxidation and in particular do not exhibit the 'breakaway' phenomenon even after a prolonged treatment at 600 deg C. (authors) [fr

  3. 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.

  4. 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.

  5. 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)

  6. 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

  7. 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)

  8. 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)

  9. 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.

  10. 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)

  11. 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

  12. Management of waste cladding hulls. Part II. An assessment of zirconium pyrophoricity and recommendations for handling waste hulls

    International Nuclear Information System (INIS)

    Kullen, B.J.; Levitz, N.M.; Steindler, M.J.

    1977-11-01

    This report reviews experience and research related to the pyrophoricity of zirconium and zirconium alloys. The results of recent investigations of the behavior of Zircaloy and some observations of industrial handling and treatment of Zircaloy tubing and scrap are also discussed. A model for the management of waste Zircaloy cladding hulls from light water reactor fuel reprocessing is offered, based on an evaluation of the reviewed information. It is concluded that waste Zircaloy cladding hulls do not constitute a pyrophoric hazard if, following the model flow sheet, finely divided metal is oxidized during the management procedure. Steps alternative to the model are described which yield zirconium in deactivated form and also accomplish varying degrees of transuranic decontamination. Information collected into appendixes is (1) a collation of zirconium pyrophoricity data from the literature, (2) calculated radioactivity contents in Zircaloy cladding hulls from spent LWR fuels, and (3) results of a laboratory study on volatilization of zirconium from Zircaloy using HCl or Cl 2

  13. Method of and apparatus for use in joining tubular components and tube assemblies made thereby

    International Nuclear Information System (INIS)

    Percival, S.R.

    1979-01-01

    A method of joining difficult to weld materials involves the forming of a rolled joint. A particular application is the joining of zirconium alloy calandria tubes to stainless steel tube-plates in a SGHWR. (UK)

  14. 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)

  15. Zirconium and hafnium

    Science.gov (United States)

    Jones, James V.; Piatak, Nadine M.; Bedinger, George M.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Zirconium and hafnium are corrosion-resistant metals that are widely used in the chemical and nuclear industries. Most zirconium is consumed in the form of the main ore mineral zircon (ZrSiO4, or as zirconium oxide or other zirconium chemicals. Zirconium and hafnium are both refractory lithophile elements that have nearly identical charge, ionic radii, and ionic potentials. As a result, their geochemical behavior is generally similar. Both elements are classified as incompatible because they have physical and crystallochemical properties that exclude them from the crystal lattices of most rock-forming minerals. Zircon and another, less common, ore mineral, baddeleyite (ZrO2), form primarily as accessory minerals in igneous rocks. The presence and abundance of these ore minerals in igneous rocks are largely controlled by the element concentrations in the magma source and by the processes of melt generation and evolution. The world’s largest primary deposits of zirconium and hafnium are associated with alkaline igneous rocks, and, in one locality on the Kola Peninsula of Murmanskaya Oblast, Russia, baddeleyite is recovered as a byproduct of apatite and magnetite mining. Otherwise, there are few primary igneous deposits of zirconium- and hafnium-bearing minerals with economic value at present. The main ore deposits worldwide are heavy-mineral sands produced by the weathering and erosion of preexisting rocks and the concentration of zircon and other economically important heavy minerals, such as ilmenite and rutile (for titanium), chromite (for chromium), and monazite (for rare-earth elements) in sedimentary systems, particularly in coastal environments. In coastal deposits, heavy-mineral enrichment occurs where sediment is repeatedly reworked by wind, waves, currents, and tidal processes. The resulting heavy-mineral-sand deposits, called placers or paleoplacers, preferentially form at relatively low latitudes on passive continental margins and supply 100 percent of

  16. Corrosion resistant zirconium alloys prepared by powder metallurgy

    International Nuclear Information System (INIS)

    Wojeik, C.C.

    1984-01-01

    Pure zirconium and zirconium 2.5% niobium were prepared by powder metallurgy. The powders were prepared directly from sponge and consolidated by cold isostatic pressing and sintering. Hot isostatic pressing was also used to obtain full density after sintering. For pure zirconium the effects of particle size, compaction pressure, sintering temperature and purity were investigated. Fully densified zirconium and Zr-2.5%Nb exhibited tensile properties comparable to cast material at room temperature and 300 0 F (149 0 C). Pressed and sintered material having density of 94-99% had slightly lower tensile properties. Corrosion tests were performed in boiling 65% H/sub 2/SO/sub 4/, 70% HNO/sub 3/, 20% HCl and 20% HCl + 500 ppm FeCl/sub 3/ (a known pitting solution). For fully dense material the observed corrosion behavior was nearly equivalent to cast material. A slightly higher rate of attack was observed for samples which were only 94-99% dense. Welding tests were also performed on zirconium and Zr-2.5%Nb alloy. Unlike P/M titanium alloys, these materials had good weldability due to the lower content of volatile impurities in the powder. A slight amount of weld porosity was observed but joint efficiencies were always not 100%, even for 94-99% density samples. Several practical applications of the P/M processed material will be briefly described

  17. Production kinetics of zirconium tetrachloride

    International Nuclear Information System (INIS)

    Sudjoko, D.; Masduki, B.; Sunardjo; Sulistyo, B.

    1996-01-01

    This research was intended to study the kinetics of zirconium tetrachloride production. The process was carried out in semi continuous reactor, equipped with heater, temperature controller, sublimator and scrubber. The variables investigated were time, temperature and the pellet forming pressure. Within the range of variables studied, the expression of the process in the chemical reaction controller region and diffusion controller region were both presented. (author)

  18. 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)

  19. 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%.

  20. 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.

  1. High temperature evaporation of titanium, zirconium and hafnium carbides

    International Nuclear Information System (INIS)

    Gusev, A.I.; Rempel', A.A.

    1991-01-01

    Evaporation of cubic nonstoichiometric carbides of titanium, zirconium and hafnium in a comparatively low-temperature interval (1800-2700) with detailed crystallochemical sample certification is studied. Titanium carbide is characterized by the maximum evaporation rate: at T>2300 K it loses 3% of sample mass during an hour and at T>2400 K titanium carbide evaporation becomes extremely rapid. Zirconium and hafnium carbide evaporation rates are several times lower than titanium carbide evaporation rates at similar temperatures. Partial pressures of metals and carbon over the carbides studied are calculated on the base of evaporation rates

  2. 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)

  3. 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)

  4. 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.

  5. Radiochemical neutron activation analysis of zirconium and zirconium-niobium alloys

    International Nuclear Information System (INIS)

    Tashimova, F.A.; Sadikov, I.I.; Salimov, M.

    2004-01-01

    Full text: Zirconium and zirconium-niobium alloys are used on nuclear technology, as fuel cladding of nuclear reactors. Their nuclear-physical, mechanical and thermophysical properties are influenced them matrix and impurity composition, therefore determination of matrix and impurity content of these materials is a very important task. Neutron activation analysis is one from multielemental and high sensible techniques that are widely applied in analysis of high purity materials. Investigation of nuclear-physical characteristics of zirconium has shown that instrumental variant NAA is unusable for analysis due to high radioactivity of a matrix. Therefore it is necessary carrying out radiochemical separation of impurity radionuclides from matrix. Study of the literature datum have shown, that zirconium and niobium are very well extracted from muriatic solution with 5% tributyl phosphineoxide (TBPO) solution in toluene and 0,75 M solution of di-2-ethyl hexyl phosphoric acid (HDEHP) in cyclohexanone. Investigation of these elements extraction in these systems has shown that more effective and selective separation of matrix radionuclides is achieved in HDEHP-3M HCI system. This system is also extracted and hafnium, witch is an accompanying element of zirconium and its high content prevented determination of other impurity elements in sample. Therefore we used extraction system HDEHP-3M HCl for analysis of zirconium and zirconium-niobium alloys in chromatographic variant. By measurement of distribution profile of a matrix and of elution curve of determined elements is established, that for effective separation of impurity and matrix radionuclides there is enough chromatographic column with diameter 1 cm and height of a sorbent layer 7 cm, thus volume of elute, necessary for complete elution of determinate elements is 35-40 ml. On the basis of the carried out researches the technique of radiochemical NAA of high purity zirconium and zirconium-niobium alloy, which allows to

  6. History of the development of zirconium alloys for use in nuclear reactors

    International Nuclear Information System (INIS)

    Rickover, H.G.; Geiger, L.D.; Lustman, B.

    1975-01-01

    The technical problems and the major decisions made during the early development of zirconium alloys for use in naval reactors are outlined. A summary is given of the development of commercial sources of supply for zirconium and hafnium metal over the period 1950 to 1965, and the problems encountered in obtaining zirconium needed for early naval prototype and shipboard reactors are identified. Steps taken in the Government procurement process are described and statistics on production amounts, prices, and inventory are included. Also included are the technical aspects associated with the development of zirconium for water-cooled nuclear reactors, beginning in early 1949 when the Bettis Atomic Power Laboratory was established as a part of the Naval Reactors Program. While in the course of the next 25 years, small-scale investigations were performed on other potential core structural materials such as stainless steel, niobium, aluminum, and beryllium, the pressure for continual development, improvement, and application of zirconium was predominant and unrelenting. (U.S.)

  7. 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

  8. Zirconium metal-water oxidation kinetics. V. Oxidation of Zircaloy in high pressure steam

    International Nuclear Information System (INIS)

    Pawel, R.E.; Cathcart, J.V.; Campbell, J.J.; Jury, S.H.

    1977-12-01

    A series of scoping tests to determine the influence of steam pressure on the isothermal oxidation kinetics of Zircaloy-4 PWR tubing was undertaken. The oxidation experiments were conducted in flowing steam at 3.45, 6.90, and 10.34 MPa (500, 1000, and 1500 psi) at 905 0 C (1661 0 F), and at 3.45 and 6.90 MPa at 1101 0 C (2014 0 F). A comparison of the results of these experiments with those obtained for oxidation in steam at atmospheric pressure under similar conditions indicated that measurable enhancement of the oxidation rate occurred with increasing pressure at 905 0 C, but not at 1100 0 C

  9. 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)

  10. 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

  11. 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)

  12. Method of reducing zirconium

    International Nuclear Information System (INIS)

    Megy, J.A.

    1980-01-01

    A method was developed for making nuclear-grade zirconium from a zirconium compound, which ismore economical than previous methods since it uses aluminum as the reductant metal rather than the more expensive magnesium. A fused salt phase containing the zirconium compound to be reduced is first prepared. The fused salt phase is then contacted with a molten metal phase which contains aluminum and zinc. The reduction is effected by mutual displacment. Aluminum is transported from the molten metal phase to the fused salt phase, replacing zirconium in the salt. Zirconium is transported from the fused salt phase to the molten metal phase. The fused salt phase and the molten metal phase are then separated, and the solvent metal and zirconium are separated by distillation or other means. (DN)

  13. Effect of high hydrogen content on metallurgical and mechanical properties of zirconium alloy claddings after heat-treatment at high temperature

    International Nuclear Information System (INIS)

    Turque, Isabelle

    2016-01-01

    Under hypothetical loss-of-coolant accident conditions, fuel cladding tubes made of zirconium alloys can be exposed to steam at high temperature (HT, up 1200 C) before being cooled and then quenched in water. In some conditions, after burst occurrence the cladding can rapidly absorb a significant amount of hydrogen (secondary hydriding), up to 3000 wt.ppm locally, during steam exposition at HT. The study deals with the effect, poorly studied up to date, of high contents of hydrogen on the metallurgical and mechanical properties of two zirconium alloys, Zircaloy-4 and M5, during and after cooling from high temperatures, at which zirconium is in its β phase. A specific facility was developed to homogeneously charge in hydrogen up to ∼ 3000 wt.ppm cladding tube samples of several centimeters in length. Phase transformations, chemical element partitioning and hydrogen precipitation during cooling from the β temperature domain of zirconium were studied by using several techniques, for the materials containing up to ∼ 3000 wt.ppm of hydrogen in average: in-situ neutron diffraction upon cooling from 700 C, X-ray diffraction, μ-ERDA, EPMA and electron microscopy in particular. The results were compared to thermodynamic predictions. In order to study the effect of high hydrogen contents on the mechanical behavior of the (prior-)μ phase of zirconium, axial tensile tests were performed at various temperatures between 20 and 700 C upon cooling from the β temperature domain, on samples with mean hydrogen contents up to ∼ 3000 wt.ppm. The results show that metallurgical and mechanical properties of the (prior-)β phase of zirconium alloys strongly depend on temperature and hydrogen content. (author) [fr

  14. 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)

  15. Method for automatic filling of nuclear fuel rod cladding tubes

    International Nuclear Information System (INIS)

    Bezold, H.

    1979-01-01

    Prior to welding the zirconium alloy cladding tubes with end caps, they are automatically filled with nuclear fuel tablets and ceramic insulating tablets. The tablets are introduced into magazine drums and led through a drying oven to a discharging station. The empty cladding tubes are removed from this discharging station and filled with tablets. A filling stamp pushes out the columns of tablets in the magazine tubes of the magazine drum into the cladding tube. Weight and measurement of length determine the filled state of the cladding tube. The cladding tubes are then led to the welding station via a conveyor belt. (DG) [de

  16. 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.

  17. PLUTONIUM-ZIRCONIUM ALLOYS

    Science.gov (United States)

    Schonfeld, F.W.; Waber, J.T.

    1960-08-30

    A series of nuclear reactor fuel alloys consisting of from about 5 to about 50 at.% zirconium (or higher zirconium alloys such as Zircaloy), balance plutonium, and having the structural composition of a plutonium are described. Zirconium is a satisfactory diluent because it alloys readily with plutonium and has desirable nuclear properties. Additional advantages are corrosion resistance, excellent fabrication propenties, an isotropie structure, and initial softness.

  18. Development of heat treated Zr-2.5 Wt% Nb pressure tube and its microstructural characterization using electron microscopy techniques

    International Nuclear Information System (INIS)

    Saibaba, N.

    2010-01-01

    Two phase Zr-2.5 wt % Nb alloy is widely used for manufacture of pressure tubes for pressurized heavy water reactors (PHWRs). These tubes are used in cold worked and stress relieved (CWSRs) condition and are manufactured by cold drawing or pilgering routes. The microstructure of the CWSR tube is characterized with presence of discontinuous β phase stringers sandwiched between elongated α-phase. Pressure tube undergoes dimensional changes and micro structural deterioration under the reactor operating conditions of temperature, pressure and neutron flux. This limits the life of the component and the availability of the power reactors. There is renewed interest in increasing the life of the pressure tube by bringing about a change in the microstructure of Zr-2.5 Nb material using various thermo mechanical processes during its manufacturing. Heat treatment of this two-phase alloy has been understood to uniquely stabilize the microstructure, which prevents degradation, under in-reactor service condition. This paper illustrates various heat treatment cycles carried out at intermediate cold working stage. Heat treatment involves solutionization of the Zr-2.5 wt % Nb tube from different temperatures followed by two types of quenching process viz, gas quenching and water quenching. The OIM-TEM studies were carried out for characterization of final tube. The technique confirmed the presence of β-phase relatively enriched in Nb content. The resulting SEM microstructures after ageing treatment at different soaking temperatures and time have been presented. Mechanical properties of heat treated pressure tubes, both at room temperature and elevated temperature have been compared with conventional CWSR pressure tube used in PHWRs. (author)

  19. 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

  20. 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)

  1. 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.

  2. Unloading Effect on Delayed Hydride Cracking in Zirconium Alloys

    International Nuclear Information System (INIS)

    Kim, Young Suk; Kim, Sung Soo

    2010-01-01

    It is well-known that a tensile overload retards not only the crack growth rate (CGR) in zirconium alloys during the delayed hydride cracking (DHC) tests but also the fatigue crack growth rate in metals, the cause of which is unclear to date. A considerable decrease in the fatigue crack growth rate due to overload is suggested to occur due either to the crack closure or to compressive stresses or strains arising from unloading of the overload. However, the role of the crack closure or the compressive stress in the crack growth rate remains yet to be understood because of incomplete understanding of crack growth kinetics. The aim of this study is to resolve the effect of unloading on the CGR of zirconium alloys, which comes in last among the unresolved issues as listed above. To this end, the CGRs of the Zr-2.5Nb tubes were determined at a constant temperature under the cyclic load with the load ratio, R changing from 0.13 to 0.66 where the extent of unloading became higher at the lower R. More direct evidence for the effect of unloading after an overload is provided using Simpson's experiment investigating the effect on the CGR of a Zr-2.5Nb tube of the stress states of the prefatigue crack tip by unloading or annealing after the formation of a pre-fatigue crack

  3. ZIRCONIUM-CLADDING OF THORIUM

    Science.gov (United States)

    Beaver, R.J.

    1961-11-21

    A method of cladding thorium with zirconium is described. The quality of the bond achieved between thorium and zirconium by hot-rolling is improved by inserting and melting a thorium-zirconium alloy foil between the two materials prior to rolling. (AEC)

  4. Spectroscopic analysis of Zirconium plasma in different ambient and optimizing conditions for nanoclusters formation

    International Nuclear Information System (INIS)

    Yadav, Dheerendra; Thareja, Raj K.

    2010-01-01

    The laser produced zirconium plasma has been studied by emission spectroscopy and fast photography using intensified charged coupled device at different ambient pressures of nitrogen (0.1, 1.0 and 10 mbar). Formation of zirconium clusters are arising at ambient pressure of 1.0 mbar at the plume periphery due to the chemical reactions between the plasma plume and the ambient and confirmed using optical emission spectroscopy. The optimum parameters for existence cluster formation are reported. The ZrN clusters are deposited on silicon substrate and characterized by AFM, XRD and EDAX techniques. (author)

  5. High energy beam thermal processing of alpha zirconium alloys and the resulting articles

    International Nuclear Information System (INIS)

    Sabol, G.P.; McDonald, S.G.; Nurminen, J.I.

    1983-01-01

    Alpha zirconium alloy fabrication methods and resultant products exhibiting improved high temperature, high pressure steam corrosion resistance. The process, according to one aspect of this invention, utilizes a high energy beam thermal treatment to provide a layer of beta treated microstructure on an alpha zirconium alloy intermediate product. The treated product is then alpha worked to final size. According to another aspect of the invention, high energy beam thermal treatment is used to produce an alpha annealed microstructure in a Zircaloy alloy intermediate size or final size component. The resultant products are suitable for use in pressurized water and boiling water reactors

  6. Process and equipement for zone heat treatment of zirconium alloys tubes

    International Nuclear Information System (INIS)

    Kiesler, A.J.; Frischmann, P.G.; Rockwood, A.C.

    1977-01-01

    Process for the thermal treatment of an area of a long zirconium alloy part in order to increase its corrosion resistance in the cooling conditions of boiling water reactor, in which the part is moved lengthwise through a succession of critical maximum temperature areas, during a critical time and is subjected to a temperature reduction at critical rate, so that each successive portion reaches a maximum temperature between 825 0 C, and directing water at a temperature around 60 to 80 0 C as jets in the cooling area [fr

  7. 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.

  8. 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

  9. The transformation behaviour of the beta phase in Zr-2.5 wt% Nb pressure tubes

    International Nuclear Information System (INIS)

    Griffiths, M.; Winegar, J.E.

    1994-12-01

    A temperature-time-transformation (TTT) diagram has been developed for the β-phase in Zr-2.5 wt% Nb pressure tubes. The results show that the morphology and/or physical state of the β-phase in pressure tubes has a significant effect on the transformation behaviour compared with a bulk Zr-19 wt%Nb alloy. (author). 14 refs., 1 tab., 15 figs

  10. 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

  11. Delayed hydride cracking behavior of Zr-2.5Nb alloy pressure tubes for PHWR700

    Energy Technology Data Exchange (ETDEWEB)

    Sunil, S.; Bind, A.K.; Khandelwal, H.K.; Singh, R.N., E-mail: rnsingh@barc.gov.in; Chakravartty, J.K.

    2015-11-15

    In order to attain improved in-reactor performance few prototypes pressure tubes of Zr-2.5Nb alloy were manufactured by employing forging to break the cast structure and to obtain more homogeneous microstructure. Both double forging and single forging were employed. The forged material was further processed by employing hot extrusion, cold pilgering and autoclaving. A detailed characterization in terms of mechanical properties and microstructure of the prototype tubes were carried for qualifying it for intended use as pressure tubes in PHWR700 reactors. In this work, Delayed Hydride Cracking (DHC) behavior of the forged Zr-2.5Nb pressure tube material characterized in terms of DHC velocity and threshold stress intensity factor associated with DHC (K{sub IH}) was compared with that of conventionally manufactured material in the temperature range of 200–283 °C. Activation energy associated with the DHC in this alloy was found to be ∼60 kJ/mol for the forged materials.

  12. 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)

  13. 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)

  14. Metallurgy of zirconium and hafnium

    International Nuclear Information System (INIS)

    Baryshnikov, N.V.; Geger, V.Eh.; Denisova, N.D.; Kazajn, A.A.; Kozhemyakin, V.A.; Nekhamkin, L.G.; Rodyakin, V.V.; Tsylov, Yu.A.

    1979-01-01

    Considered are those properties of zirconium and of hafnium, which are of practical interest for the manufacture of these elements. Systematized are the theoretical and the practical data on the procedures for thermal decomposition of zirconia and for obtaining zirconium dioxide and hafnium dioxide by a thermal decomposition of compounds and on the hydrometallurgical methods for extracting zirconium and hafnium. Zirconium and hafnium fluorides and chlorides production procedures are described. Considered are the iodide and the electrolytic methods of refining zirconium and hafnium

  15. 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.

  16. Synthesis of zirconium by zirconium tetrachloride reduction by magnesio-thermia. Experimental study and modelling; Elaboration de zirconium par reduction de tetrachlorure de zirconium par magnesothermie. Etude experimentale et modelisation

    Energy Technology Data Exchange (ETDEWEB)

    Basin, N

    2001-01-01

    This work deals with the synthesis of zirconium. The ore is carbo-chlorinated to obtain the tetrachloride which is then purified by selective condensation and extractive distillation. Zirconium tetrachloride is then reduced by magnesium and the pseudo-alloy is obtained according to the global following reaction (Kroll process): ZrCl{sub 4} + 2 Mg = 2 MgCl{sub 2}. By thermodynamics, it has been shown that the volatilization of magnesium chloride and the formation of zirconium sub-chlorides are minimized when the combined effects of temperature and of dilution with argon are limited. With these conditions, the products, essentially zirconium and magnesium chloride, are obtained in equivalence ratio in the magnesio-thermia reaction. The global kinetics of the reduction process has been studied by a thermal gravimetric method. A thermo-balance device has been developed specially for this kinetics study. It runs under a controlled atmosphere and is coupled to a vapor tetrachloride feed unit. The transformation is modelled supposing that the zirconium and magnesium chloride formation result: 1)of the evaporation of magnesium from its liquid phase 2)of the transfer of magnesium and zirconium tetrachloride vapors towards the front of the reaction located in the gaseous phase 3)of the chemical reaction. In the studied conditions, the diffusion is supposed to be the limiting process. The influence of the following parameters: geometry of the reactive zone, temperature, scanning rate of the argon-zirconium tetrachloride mixture, composition of the argon-zirconium tetrachloride mixture has been experimentally studied and confronted with success to the model. (O.M.)

  17. 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)

  18. 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)

  19. 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)

  20. 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.

  1. 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)

  2. Contribution to the study of zirconium self-diffusion in zirconium carbide

    International Nuclear Information System (INIS)

    An, Chul

    1972-01-01

    The objective of this research thesis is to determine experimental conditions allowing the measurement of the self-diffusion coefficient of zirconium in zirconium carbide. The author reports the development of a method of preparation of zirconium carbide samples. He reports the use of ion implantation as technique to obtain a radio-tracer coating. The obtained results give evidence of the impossibility to use sintered samples with small grains because of the demonstrated importance of intergranular diffusion. The self-diffusion coefficient is obtained in the case of zirconium carbide with grains having a diameter of few millimetres. The presence of 95 Nb from the disintegration of 95 Zr indicates that these both metallic elements have very close diffusion coefficients at 2.600 C [fr

  3. 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

  4. Thin-walled large-diameter zirconium alloy tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Price, E.G.; Richinson, P.J.

    1978-08-01

    The requirements of the thin-walled large-diameter Zircaloy-2 tubing used in CANDU reactors are reviewed. Strength, residual stress patterns, texture and prior deformation contribute to the stability of these tubes. The extent to which the present manufacturing route meets these requirements is discussed. (author)

  5. New zirconium alloys for nuclear application; Novas ligas de zirconio para aplicacao nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Lobo, R.M.; Andrade, A.H.P., E-mail: rmlobo@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2010-07-01

    Zirconium alloys are widely used in the nuclear industry, mainly in fuel cladding tubes and structural components for PWR plants. The service life of these components, which operate under high temperatures conditions ({approx} 300 deg C), has led to developing new alloys with the aim to improve the mechanical properties, corrosion resistance and irradiation damage. The variation in the composition of the alloy produces second phase particles which alter the materials properties according to their size and distribution, is essential therefore, knowledge their characteristics. Analysis of second phase particles in zirconium alloys are carried out by scanning electron microscopy, transmission electron microscopy and image analysis. This study used the zircaloy-4 to illustrate the characterization of these alloys through the study of second phase particles. (author)

  6. 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)

  7. 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

  8. 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)

  9. 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)

  10. 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.

  11. 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

  12. Artefacts in multimodal imaging of titanium, zirconium and binary titanium-zirconium alloy dental implants: an in vitro study.

    Science.gov (United States)

    Smeets, Ralf; Schöllchen, Maximilian; Gauer, Tobias; Aarabi, Ghazal; Assaf, Alexandre T; Rendenbach, Carsten; Beck-Broichsitter, Benedicta; Semmusch, Jan; Sedlacik, Jan; Heiland, Max; Fiehler, Jens; Siemonsen, Susanne

    2017-02-01

    To analyze and evaluate imaging artefacts induced by zirconium, titanium and titanium-zirconium alloy dental implants. Zirconium, titanium and titanium-zirconium alloy implants were embedded in gelatin and MRI, CT and CBCT were performed. Standard protocols were used for each modality. For MRI, line-distance profiles were plotted to quantify the accuracy of size determination. For CT and CBCT, six shells surrounding the implant were defined every 0.5 cm from the implant surface and histogram parameters were determined for each shell. While titanium and titanium-zirconium alloy induced extensive signal voids in MRI owing to strong susceptibility, zirconium implants were clearly definable with only minor distortion artefacts. For titanium and titanium-zirconium alloy, the MR signal was attenuated up to 14.1 mm from the implant. In CT, titanium and titanium-zirconium alloy resulted in less streak artefacts in comparison with zirconium. In CBCT, titanium-zirconium alloy induced more severe artefacts than zirconium and titanium. MRI allows for an excellent image contrast and limited artefacts in patients with zirconium implants. CT and CBCT examinations are less affected by artefacts from titanium and titanium-zirconium alloy implants compared with MRI. The knowledge about differences of artefacts through different implant materials and image modalities might help support clinical decisions for the choice of implant material or imaging device in the clinical setting.

  13. Artefacts in multimodal imaging of titanium, zirconium and binary titanium–zirconium alloy dental implants: an in vitro study

    Science.gov (United States)

    Schöllchen, Maximilian; Aarabi, Ghazal; Assaf, Alexandre T; Rendenbach, Carsten; Beck-Broichsitter, Benedicta; Semmusch, Jan; Sedlacik, Jan; Heiland, Max; Fiehler, Jens; Siemonsen, Susanne

    2017-01-01

    Objectives: To analyze and evaluate imaging artefacts induced by zirconium, titanium and titanium–zirconium alloy dental implants. Methods: Zirconium, titanium and titanium–zirconium alloy implants were embedded in gelatin and MRI, CT and CBCT were performed. Standard protocols were used for each modality. For MRI, line–distance profiles were plotted to quantify the accuracy of size determination. For CT and CBCT, six shells surrounding the implant were defined every 0.5 cm from the implant surface and histogram parameters were determined for each shell. Results: While titanium and titanium–zirconium alloy induced extensive signal voids in MRI owing to strong susceptibility, zirconium implants were clearly definable with only minor distortion artefacts. For titanium and titanium–zirconium alloy, the MR signal was attenuated up to 14.1 mm from the implant. In CT, titanium and titanium–zirconium alloy resulted in less streak artefacts in comparison with zirconium. In CBCT, titanium–zirconium alloy induced more severe artefacts than zirconium and titanium. Conclusions: MRI allows for an excellent image contrast and limited artefacts in patients with zirconium implants. CT and CBCT examinations are less affected by artefacts from titanium and titanium–zirconium alloy implants compared with MRI. The knowledge about differences of artefacts through different implant materials and image modalities might help support clinical decisions for the choice of implant material or imaging device in the clinical setting. PMID:27910719

  14. 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.

  15. 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

  16. Method of separating hafnium from zirconium

    International Nuclear Information System (INIS)

    Megy, J.A.

    1980-01-01

    English. A new anhydrous method was developed for separating zirconium and hafnium, which gives higher separation factors and is more economical than previous methods. A molten phase, comprising a solution of unseparated zirconium and hafnium and a solvent metal, is first prepared. The molten metal phase is contacted with a fused salt phase which includes a zirconium salt. Zirconium and hafnium separation is effected by mutual displacement with hafnium being transported from the molten metal phase to the fused salt phase, while zirconium is transported from the fused salt phase to the molten metal phase. The solvent metal is less electropositive than zirconium. Zinc was chosen as the solvent metal, from a group which also included cadmium, lead, bismuth, copper, and tin. The fused salt phase cations are more electropositive than zirconium and were selected from a group comprising the alkali elements, the alkaline earth elements, the rare earth elements, and aluminum. A portion of the zirconium in the molten metal phase was oxidized by injecting an oxidizing agent, chlorine, to form zirconium tetrachlorid

  17. Ab initio atomic simulation of hydrogen and iodine effects in zirconium

    International Nuclear Information System (INIS)

    Domain, Ch.

    2002-03-01

    In this work we present ab initio atomic simulations concerning the effects of hydrogen and iodine in hexagonal zirconium. We first studied the point defects in the dilute Zr-H (and to a less extend Zr-H-O) systems and concluded that it is better described within the generalised gradient approximation for the exchange and correlation functional. We calculated the hydrogen thermal diffusion coefficient in solid solution that agree very well with the experimental values. The calculated formation energy of different self-interstitial configuration are rather small (around 3 eV) and close to each other indicating the high complexity of these defects. We studied the core structure of the screw dislocation that has a preferential prismatic spreading. We also calculated the gamma surface for different gliding planes. The influence of hydrogen, that induces a significant reduction of the gamma surfaces excess energies, allows to qualitatively explain experimental results regarding some hydrogen effects on hexagonal zirconium plastic deformation. We also discussed the effect of zirconium hydride stoichiometry on gamma surfaces. The results concerning the iodine and oxygen adsorption on zirconium surfaces, inducing the evaluation of the effective surface energy reduction as a function of the iodine partial pressure allow for a better description of iodine induced stress corrosion cracking of zirconium. (author)

  18. 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)

  19. 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)

  20. 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.

  1. Characterization of magnetically impelled arc butt welded T11 tubes for high pressure applications

    Directory of Open Access Journals (Sweden)

    R. Sivasankari

    2015-09-01

    Full Text Available Magnetically impelled arc butt (MIAB welding is a pressure welding process used for joining of pipes and tubes with an external magnetic field affecting arc rotation along the tube circumference. In this work, MIAB welding of low alloy steel (T11 tubes were carried out to study the microstructural changes occurring in thermo-mechanically affected zone (TMAZ. To qualify the process for the welding applications where pressure could be up to 300 bar, the MIAB welds are studied with variations of arc current and arc rotation time. It is found that TMAZ shows higher hardness than that in base metal and displays higher weld tensile strength and ductility due to bainitic transformation. The effect of arc current on the weld interface is also detailed and is found to be defect free at higher values of arc currents. The results reveal that MIAB welded samples exhibits good structural property correlation for high pressure applications with an added benefit of enhanced productivity at lower cost. The study will enable the use of MIAB welding for high pressure applications in power and defence sectors.

  2. Zirconium-barrier cladding attributes

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.; Rand, R.A.; Tucker, R.P.; Cheng, B.; Adamson, R.B.; Davies, J.H.; Armijo, J.S.; Wisner, S.B.

    1987-01-01

    This metallurgical study of Zr-barrier fuel cladding evaluates the importance of three salient attributes: (1) metallurgical bond between the zirconium liner and the Zircaloy substrate, (2) liner thickness (roughly 10% of the total cladding wall), and (3) softness (purity). The effect that each of these attributes has on the pellet-cladding interaction (PCI) resistance of the Zr-barrier fuel was studied by a combination of analytical model calculations and laboratory experiments using an expanding mandrel technique. Each of the attributes is shown to contribute to PCI resistance. The effect of the zirconium liner on fuel behavior during off-normal events in which steam comes in contact with the zirconium surface was studied experimentally. Simulations of loss-of-coolant accident (LOCA) showed that the behavior of Zr-barrier cladding is virtually indistinguishable from that of conventional Zircaloy cladding. If steam contacts the zirconium liner surface through a cladding perforation and the fuel rod is operated under normal power conditions, the zirconium liner is oxidized more rapidly than is Zircaloy, but the oxidation rate returns to the rate of Zircaloy oxidation when the oxide phase reaches the zirconium-Zircaloy metallurgical bond

  3. Experimental studies of relevance on zirconium nitrate raffinate sludge for its disposal as well as zirconium recovery

    International Nuclear Information System (INIS)

    Brahmananda Reddy, G.; Narasimha Murty, B.; Ravindra, H.R.

    2013-01-01

    One of the many routes of production of nuclear grade zirconium dioxide involve separation of zirconium and hafnium by solvent extraction of zirconium nitrate using tri-n-butyl phosphate followed by precipitation of zirconium with ammonia and finally calcination of the so obtained hydrated zirconia at elevated temperature. The zirconium feed solution as is generated from digestion of zirconium washed dried frit (produced by the caustic fusion of zircon sand which is one of the beach sand heavy minerals) in nitric acid contain considerable amount of sludge material and after solvent extraction this whole sludge material rests with raffinate. This sludge material has a scope to contain considerable amounts of zirconium along with other metal ions such as hafnium, aluminium, iron, etc. besides nitric acid and it constitutes one of the important solid wastes that needs to be disposed suitably. One of the disposal means of this sludge material is to use it as a land fill for which two important criteria are to be viz the pH of 10% solid waste solution should be near to neutral pH and the loss on ignition at 550℃ on dry basis of the sludge to be below 20%. In order to study the implications of presence of varying amounts of zirconium nitrate in the sludge on the pH of 10% solution of the sludge various synthetic zirconium nitrate solid waste were prepared using the sludge material generated at the laboratory during the analysis of zirconium washed dried frit. Presence of zirconium in the sludge is expected to decrease the overall pH of the 10% solution of the sludge because zirconium is prone to hydrolyze especially locally when zirconium ion comes into contact with water according to the chemical equation Zr 4+ H 2 O → ZrO 2+ + 2H + . From this equation, it is clear that for every one mole of zirconium ions two moles of hydrogen ions are produced. This is verified experimentally using the synthetically prepared sludge materials with varying amounts of zirconium

  4. An experimental study on two-phase pressure drop in small diameter horizontal, downward inclined and vertical tubes

    Directory of Open Access Journals (Sweden)

    Autee Arun

    2015-01-01

    Full Text Available An experimental study of two-phase pressure drop in small diameter tubes orientated horizontally, vertically and at two other downward inclinations of θ= 300 and θ = 600 is described in this paper. Acrylic transparent tubes of internal diameters 4.0, 6.0, and 8.0 mm with lengths of 400 mm were used as the test section. Air-water mixture was used as the working fluid. Two-phase pressure drop was measured and compared with the existing correlations. These correlations are commonly used for calculation of pressure drop in macro and mini-microchannels. It is observed that the existing correlations are inadequate in predicting the two-phase pressure drop in small diameter tubes. Based on the experimental data, a new correlation has been proposed for predicting the two-phase pressure drop. This correlation is developed by modification of Chisholm parameter C by incorporating different parameters. It was found that the proposed correlation predicted two-phase pressure drop at satisfactory level.

  5. A comparison of R-22, R-134a, R-410a, and R-407c condensation performance in smooth and enhanced tubes: Part 2, Pressure drop

    Energy Technology Data Exchange (ETDEWEB)

    Eckels, S J; Tesene, B A

    1999-07-01

    This paper reports pressure drops during condensation for R-22, R-134a, R-410a, and R-407c in three enhanced tubes and one smooth tube. The test tubes were a 3/8 inch outer diameter smooth tube, a 3/8 inch outer diameter microfin tube, a 5/16 inch outer diameter microfin tube, and a 5/8 inch outer diameter microfin tube. Pressure drops are reported at four mass fluxes, at two saturation temperatures, and over a range of average qualities in the test tubes. The pressure drops for R-410a were approximately 40% lower than those of R-22 in both tubes. R-407c had 10% to 20% lower pressure drops than R-22, while 134-a had slightly larger pressure drops than R-22. The microfin tube pressure drops were, on average, 40% to 80% higher than those for the smooth tube for all refrigerants. The pressure drop penalty of the microfin tube was shown to decrease with increased quality.

  6. Failure assessment and evaluation of critical crack length for a fresh Zr-2 pressure tube of an Indian PHWR

    International Nuclear Information System (INIS)

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

    1996-01-01

    Fracture analysis of Zr-2 pressure tubes having through wall axial crack was done using finite element method. The analysis was done for tubes in as received condition. During reactor operation the mechanical properties of Zr-2 undergo changes. The analysis is valid for pressure tubes of newly commissioned reactors. The main aim of the study was to determine critical crack length of pressure tubes in normal operating conditions. Elastic plastic fracture analysis was done for different crack lengths to determine applied J-integral values. Tearing modulus instability concept was used to evaluate critical crack length. One of the important parameter studied was, the effect of crack face pressure, which leaking fluid exert on the crack faces/lips of through wall axial crack. Its effect was found to be significant for pressure tubes. It increases the applied J-integral values. Approximate analytical solutions which takes into account the plasticity ahead of crack tip, are available and widely used. These formulae do not take into account the crack face pressure. Since, for the present situation the effect of crack face pressure is significant hence, detailed finite analysis was necessary. Detailed 3D finite element analysis gives an insight into the variation of J-integral values over the thickness of pressure tube. It was found that J values are maximum at the middle layer of the tube. A peak factor on J values was defined and evaluated as ratio of maximum J to average J across the thickness, crack opening area for each length was also evaluated. The knowledge of crack opening area is useful for leak before break studies. The failure assessment was also done using Central Electricity Generating Board (CEGB) R-6 method considering the ductile tearing. The reserve factors (or safety margins) for different crack lengths was evaluated using R-6 method. (author). 30 refs., 21 figs., 34 tabs

  7. 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)

  8. Pressure Measurements on a Deforming Surface in Response to an Underwater Explosion in a Water-Filled Aluminum Tube

    Directory of Open Access Journals (Sweden)

    G. Chambers

    2001-01-01

    Full Text Available Experiments have been conducted to benchmark DYSMAS computer code calculations for the dynamic interaction of water with cylindrical structures. Small explosive charges were suspended using hypodermic needle tubing inside Al tubes filled with distilled water. Pressures were measured during shock loading by tourmaline crystal, carbon resistor and ytterbium foil gages bonded to the tube using a variety of adhesives. Comparable calculated and measured pressures were obtained for the explosive charges used, with some gages surviving long enough to record results after cavitation with the tube wall.

  9. Zirconium isotope separation process

    International Nuclear Information System (INIS)

    Peterson, S.H.; Lahoda, E.J.

    1988-01-01

    A process is described for reducing the amount of zirconium 91 isotope in zirconium comprising: forming a first solution of (a) a first solvent, (b) a scavenger, and (c) a zirconium compound which is soluble in the first solvent and reacts with the scavenger when exposed to light of a wavelength of 220 to 600 nm; irradiating the first solution with light at the wavelength for a time sufficient to photoreact a disproportionate amount of the zirconium compound containing the zirconium 91 isotope with the scavenger to form a reaction product in the first solution; contacting the first solution, while effecting the irradiation, with a second solvent which is immiscible with the first solvent, which the second solvent is a preferential solvent for the reaction product relative to the first solvent, such that at least a portion of the reaction product is transferred to the second solvent to form a second solution; and separating the second solution from the first solution after the contacting

  10. Heat exchanging tube behaviour in steam generators of pressurized water reactors

    International Nuclear Information System (INIS)

    Pastor, D.; Oertel, K.

    1979-01-01

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

  11. Application of FEM analytical method for hydrogen migration behaviour in Zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Arioka, K; Ohta, H [Takasago Research and Development Center, Mitsubishi Heavy Industries Ltd, Hyogo-ken (Japan)

    1997-02-01

    It is well recognized that the hydriding behaviours of Zirconium alloys are very significant problems as a safety issues. Also, it is well known that the diffusion of hydrogen in Zirconium alloys are affected not only by concentration but also temperature gradient. But in actual component, especially heat transfer tube such as fuel rod, we can not avoid the temperature gradient in some degree. So, it is very useful to develop the computer code which can analyze the hydrogen diffusion and precipitation behaviours under temperature gradient as a function of the structure of fuel rod. For this objective, we have developed the computer code for hydrogen migration behaviour using FEM analytical methods. So, following items are presented and discussed. Analytical method and conditions; correlation between the computed and test results; application to designing studies. (author). 8 refs, 4 figs, 2 tabs.

  12. Free Piston Double Diaphragm Shock Tube

    OpenAIRE

    OGURA, Eiji; FUNABIKI, Katsushi; SATO, Shunichi; ABE, Takashi; 小倉, 栄二; 船曳, 勝之; 佐藤, 俊逸; 安部, 隆士

    1997-01-01

    A free piston double diaphragm shock tube was newly developed for generation of high Mach number shock wave. Its characteristics was investigated for various operation parameters; such as a strength of the diaphragm at the end of the comparession tube, an initial pressure of low pressure tube, an initial pressure of medium pressure tube and the volume of compression tube. Under the restriction of fixed pressures for the driver high pressure tube (32×10^5Pa) and the low pressure tube (40Pa) in...

  13. MEASUREMENT OF ENDOTRACHEAL TUBE CUFF PRESSURE IN MECHANICALLYVENTILATED PATIENTS ON ARRIVAL TO INTENSIVE CARE UNIT - A CROSS-SECTIONAL STUDY

    Directory of Open Access Journals (Sweden)

    Arun Kumar Ajjappa

    2017-04-01

    Full Text Available BACKGROUND The monitoring of Endotracheal Tube (ETT cuff pressure in intubated patients on arrival to intensive care unit is very essential. The cuff pressure must be within an optimal range of 20-30cm H2O ensuring ventilation with no complications related to cuff overinflation and underinflation. This can be measured with a cuff pressure manometer. The aim of the study is to measure the endotracheal tube cuff pressure in patients on arrival to intensive care unit and to identify prevalence of endotracheal cuff underinflation and overinflation. MATERIALS AND METHODS A cross-sectional study was done on mechanically-ventilated patients who were intubated in casualty (emergency department on arrival to intensive care unit in S.S. Institute of Medical Sciences and Research Centre, Davangere. About 50 critically-ill patients intubated with a high volume, low pressure endotracheal tube were included in the study. An analogue manometer was used to measure the endotracheal tube cuff pressure. It was compared with the recommended level. The settings of mechanical ventilation, endotracheal tube size and peak airway pressure were recorded. RESULTS It was found that the mean cuff pressure was 64.10 cm of H2O with a standard deviation of 32.049. Of the measured cuff pressures, only 2% had pressures within an optimal range (20-30cm of H2O. 88% had cuff pressures more than 30cm of H2O. The mean peak airway pressure found to be 20.50cm of H2O with a Standard Deviation (SD of 5.064. CONCLUSION This study is done to emphasise the importance of cuff pressure measurement in all mechanically-ventilated patients as cuff pressure is found to be high in most of the patients admitted to intensive care unit. Complications of overinflation and underinflation can only be prevented if the acceptable cuff pressures are achieved.

  14. Pressure tests to assess the significance of defects in boiler and superheater tubing

    International Nuclear Information System (INIS)

    Guest, J.C.; Hutchings, J.A.

    1975-01-01

    Internal pressure tests on 9 per cent Cr-1 per cent Mo steel tubing containing artificial defects demonstrated that the resultant loss of strength was less than a simple calculation based on the reduced tube thickness would suggest. Bursting tests on tubes containing longitudinal defects of varying length, depth and acuity showed notch strengthening at ambient temperature and at 550 0 C. A flow stress concept developed for simple bursting tests was shown to apply to creep conditions at 550 0 C. Results of creep and short-term bursting tests show that the length as well as the depth of the defect is an important factor affecting the life of bursting strength of the tubes. Defects less than 10 per cent of the tube thickness were found to have an insignificant effect. (author)

  15. Zirconium for nitric acid solutions

    International Nuclear Information System (INIS)

    Yau, T.L.

    1984-01-01

    The excellent corrosion resistance of zirconium in nitric acid has been known for over 30 years. Recently, there is an increasing interest in using zirconium for nitric acid services. Therefore, an extensive research effort has been carried out to achieve a better understanding of the corrosion properties of zirconium in nitric acid. Particular attention is paid to the effect of concentration, temperature, structure, solution impurities, and stress. Immersion, autoclave, U-bend, and constant strain-rate tests were used in this study. Results of this study indicate that the corrosion resistance of zirconium in nitric acid is little affected by changes in temperature and concentration, and the presence of common impurities such as seawater, sodium chloride, ferric chloride, iron, and stainless steel. Moreover, the presence of seawater, sodium chloride, ferric chloride, and stainless steel has little effect on the stress corrosion craking (SCC) susceptibility of zirconium in 70% nitric acid at room temperatures. However, zirconium could be attacked by fluoride-containing nitric acid and the vapors of chloride-containing nitric acid. Also, high sustained tensile stresses should be avoided when zirconium is used to handle 70% nitric acid at elevated temperatures or > 70% nitric acid

  16. Process for etching zirconium metallic objects

    International Nuclear Information System (INIS)

    Panson, A.J.

    1988-01-01

    In a process for etching of zirconium metallic articles formed from zirconium or a zirconium alloy, wherein the zirconium metallic article is contacted with an aqueous hydrofluoric acid-nitric acid etching bath having an initial ratio of hydrofluoric acid to nitric acid and an initial concentration of hydrofluoric and nitric acids, the improvement, is described comprising: after etching of zirconium metallic articles in the bath for a period of time such that the etching rate has diminished from an initial rate to a lesser rate, adding hydrofluoric acid and nitric acid to the exhausted bath to adjust the concentration and ratio of hydrofluoric acid to nitric acid therein to a value substantially that of the initial concentration and ratio and thereby regenerate the etching solution without removal of dissolved zirconium therefrom; and etching further zirconium metallic articles in the regenerated etching bath

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-03-15

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

  19. Heat transfer and pressure drop in a tube bank inclined with respect to the flow

    Energy Technology Data Exchange (ETDEWEB)

    Yanez Moreno, A.A.

    1985-01-01

    This research is intended to lend understanding and to quantify the heat-transfer and fluid-flow characteristics for yawed tube banks in both staggered and in-line arrays. The investigated range of yaw angle was from 90 (crossflow) to 45/sup 0/, while the freestream Reynolds number (based on the tube diameter) ranged between 7000 and 45,000. The transverse and longitudinal center-to-center distances between the tubes were S/sub T//D = S/sub L//D = 2, respectively. The heat-transfer experiments were carried out on a row-by-row basis. Pressure drop measurements were made not only upstream and downstream of the tube bank but also within it. The patterns of fluid flow adjacent to the tubes were visualized using the oil-lampblack technique. A detailed study was carried out to determine the heat-transfer characteristics of a yawed single cylinder. The yaw angle range was between 90 and 30/sup 0/, and flow visualization was also performed. The pressure measurements showed that the overall dimensionless pressure drop for the staggered array is higher than that for the in-line array for a given Reynolds number or yaw. The flow-visualization patterns showed that the boundary layer separation depends on the yaw angle. For the single cylinder, the Nusselt number varied with the yaw angle in an undulating manner and did not correlate with the Independence Principle.

  20. Zirconium and cast zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Krone, K

    1977-04-01

    A survey is given on the occurence of zirconium, production of Zr sponge and semi-finished products, on physical and mechanical properties, production of Zr cast, composition of the commercial grades and reactor grades qualities, metal cutting, welding, corrosion behavior and use.

  1. 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)

  2. Synthesis of zirconium by zirconium tetrachloride reduction by magnesio-thermia. Experimental study and modelling

    International Nuclear Information System (INIS)

    Basin, N.

    2001-01-01

    This work deals with the synthesis of zirconium. The ore is carbo-chlorinated to obtain the tetrachloride which is then purified by selective condensation and extractive distillation. Zirconium tetrachloride is then reduced by magnesium and the pseudo-alloy is obtained according to the global following reaction (Kroll process): ZrCl 4 + 2 Mg = 2 MgCl 2 . By thermodynamics, it has been shown that the volatilization of magnesium chloride and the formation of zirconium sub-chlorides are minimized when the combined effects of temperature and of dilution with argon are limited. With these conditions, the products, essentially zirconium and magnesium chloride, are obtained in equivalence ratio in the magnesio-thermia reaction. The global kinetics of the reduction process has been studied by a thermal gravimetric method. A thermo-balance device has been developed specially for this kinetics study. It runs under a controlled atmosphere and is coupled to a vapor tetrachloride feed unit. The transformation is modelled supposing that the zirconium and magnesium chloride formation result: 1)of the evaporation of magnesium from its liquid phase 2)of the transfer of magnesium and zirconium tetrachloride vapors towards the front of the reaction located in the gaseous phase 3)of the chemical reaction. In the studied conditions, the diffusion is supposed to be the limiting process. The influence of the following parameters: geometry of the reactive zone, temperature, scanning rate of the argon-zirconium tetrachloride mixture, composition of the argon-zirconium tetrachloride mixture has been experimentally studied and confronted with success to the model. (O.M.)

  3. Accounting for the residual stress effects on the creep deformation of channel tubes

    International Nuclear Information System (INIS)

    Knizhnikov, Yu.N.; Platonov, P.A.; Ul'yanov, A.I.

    1985-01-01

    The effect of the first kind residual stresses arising in the walls of the zirconium base alloy fules in the process of fabrication on the RBMK type reactor channel tube creep is investigated. Models for calculation of the reactor component creep with account for the relaxation of residual stresses distributed by the wall thickness as well as the radiation and temperature fields are developed. On the basis of the analysis of the data obtained it is concluded that the effect of the residual stresses on the RBMK channel tube deformation for a long-term operation is negligible. But for the short-term fests the results can be noticeably distorted by this factor. The role of internal stresses can also manifest when determining the deformation of radiation elongation of the zirconium base alloy samples

  4. Anisotropic thermal creep of internally pressurized Zr-2.5Nb tubes

    International Nuclear Information System (INIS)

    Li, W.; Holt, R.A.

    2010-01-01

    The anisotropy of creep of internally pressurized cold-worked Zr-2.5Nb tubes with different crystallographic textures is reported. The stress exponent n was determined to be about three at transverse stresses from 100 to 250 MPa with an activation energy of ∼99.54 kJ/mol in the temperature range 300-400 o C. The stress exponent increased to ∼6 for transverse stresses from 250 to 325 MPa. From this data an experimental regime of 350 o C and 300 MPa was established in which dislocation glide is the likely strain-producing mechanism. Creep tests were carried out under these conditions on internally pressurized Zr-2.5Nb tubes with 18 different textures. Creep strain and creep anisotropy (ratio of axial to transverse steady-state creep rate, ε . A /ε . T ) exhibited strong dependence on crystallographic textures of the Zr-2.5Nb tubes. It was found that the values of (ε . A /ε . T ) increased as the difference between the resolved faction of basal plane normals in the transverse and radial directions (f T - f R ) increases. The tubes with the strongest radial texture showed a negative axial creep strain and a negative creep rate ratio (ε . A /ε . T ) and tubes with a strong transverse texture exhibited the positive values of steady-state creep rate ratio (ε . A /ε . T ) and good creep resistance in the transverse direction. These behaviors are qualitatively similar to those observed during irradiation creep, and also to the predictions of polycrystalline models for creep in which glide is the strain-producing mechanism and prismatic slip is the dominant system. A detailed analysis of the results using polycrystalline models may assist in understanding the anisotropy of irradiation creep.

  5. 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)

  6. Neutron activation of chlorine in zirconium and zirconium alloys use of the matrix as comparator

    International Nuclear Information System (INIS)

    Cohen, I.M.; Gomez, C.D.; Mila, M.I.

    1981-01-01

    A procedure is described for neutron activation analysis of chlorine in zirconium and zirconium alloys. Calculation of chlorine concentration is performed relative to zirconium concentration in the matrix in order to minimize effects of differences in irradiation and counting geometry. Principles of the method and the results obtained are discussed. (author)

  7. Electromechanical phase transition of a dielectric elastomer tube under internal pressure of constant mass

    Directory of Open Access Journals (Sweden)

    Song Che

    2017-05-01

    Full Text Available The electromechanical phase transition for a dielectric elastomer (DE tube has been demonstrated in recent experiments, where it is found that the unbulged phase gradually changed into bulged phase. Previous theoretical works only studied the transition process under pressure control condition, which is not consistent with the real experimental condition. This paper focuses on more complex features of the electromechanical phase transition under internal pressure of constant mass. We derive the equilibrium equations and the condition for coexistent states for a DE tube under an internal pressure, a voltage through the thickness and an axial force. We find that under mass control condition the voltage needed to maintain the phase transition increases as the process proceeds. We analyze the entire process of electromechanical phase transition and find that the evolution of configurations is also different from that for pressure control condition.

  8. Mechanistic modeling of heat transfer process governing pressure tube-to-calandria tube contact and fuel channel failure

    International Nuclear Information System (INIS)

    Luxat, J.C.

    2002-01-01

    Heat transfer behaviour and phenomena associated with ballooning deformation of a pressure tube into contact with a calandria tube have been analyzed and mechanistic models have been developed to describe the heat transfer and thermal-mechanical processes. These mechanistic models are applied to analyze experiments performed in various COG funded Contact Boiling Test series. Particular attention is given in the modeling to characterization of the conditions for which fuel channel failure may occur. Mechanistic models describing the governing heat transfer and thermal-mechanical processes are presented. The technical basis for characterizing parameters of the models from the general heat transfer literature is described. The validity of the models is demonstrated by comparison with experimental data. Fuel channel integrity criteria are proposed which are based upon three necessary and sequential mechanisms: Onset of CHF and local drypatch formation at contact; sustained film boiling in the post-contact period; and creep strain to failure of the calandria tube while in sustained film boiling. (author)

  9. 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

  10. Self-consistent calculation of steady-state creep and growth in textured zirconium

    International Nuclear Information System (INIS)

    Tome, C.N.; So, C.B.; Woo, C.H.

    1993-01-01

    Irradiation creep and growth in zirconium alloys result in anisotropic dimensional changes relative to the crystallographic axis in each individual grain. Several methods have been attempted to model such dimensional changes, taking into account the development of intergranular stresses. In this paper, we compare the predictions of several such models, namely the upper-bound, the lower-bound, the isotropic K* self-consistent (analytical) and the fully self-consistent (numerical) models. For given single-crystal creep compliances and growth factors, the polycrystal compliances predicted by the upper- and lower-bound models are unreliable. The predictions of the two self-consistent approaches are usually similar. The analytical isotropic K* approach is simple to implement and can be used to estimate the creep and growth rates of the polycrystal in many cases. The numerical fully self-consistent approach should be used when an accurate prediction of polycrystal creep is required, particularly for the important case of a closed-end internally pressurized tube. In most cases, the variations in grain shape introduce only minor corrections to the behaviour of polycrystalline materials. (author)

  11. Natural convection in vertical tubes with variable properties and prescribed pressure at the end of the tube

    International Nuclear Information System (INIS)

    Almeida Rego, O.A. de; Fernandes, E.C.

    1983-01-01

    The analysis of free convection flow development in a heated vertical open tube was established in the present work for the air, with Prandtl number equal to 0.7 and for water with Prandtl numbers equal to 1.0, 2.5 and 5.0 with variable properties and prescribed pressure conditions at the end of the tube. It is considered that the flow is incompressible, laminar and stable and can be described by the continuity, momentum and energy equations with the usual boundary-layer assumptions. The equations were solved by finite difference method and from the velocity and temperature distributions many quantities such as dimensioless flow and heat rates and Nusselt numbers can be determined. (Author) [pt

  12. 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

  13. Eccentric pressurized tube for measuring creep rupture

    International Nuclear Information System (INIS)

    Schwab, P.R.

    1981-01-01

    Creep rupture is a long term failure mode in structural materials that occurs at high temperatures and moderate stress levels. The deterioration of the material preceding rupture, termed creep damage, manifests itself in the formation of small cavities on grain boundaries. To measure creep damage, sometimes uniaxial tests are performed, sometimes density measurements are made, and sometimes the grain boundary cavities are measured by microscopy techniques. The purpose of the present research is to explore a new method of measuring creep rupture, which involves measuring the curvature of eccentric pressurized tubes. Theoretical investigations as well as the design, construction, and operation of an experimental apparatus are included in this research

  14. 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

  15. 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

  16. Peculiarities of formation of zirconium aluminides in hydride cycle mode

    International Nuclear Information System (INIS)

    Muradyan, G.N.

    2016-01-01

    The zirconium aluminides are promising structural materials in aerospace, mechanical engineering, chemical industry, etc. They are promising for manufacturing of heat-resistant wires, that will improve the reliability and efficiency of electrical networks. In the present work, the results of study of zirconium aluminides formation in the Hydride Cycle (HC) mode, developed in the Laboratory of high-temperature synthesis of the Institute of Chemical Physics of NAS RA, are described. The formation of zirconium aluminides in HC proceeded according to the reaction xZrH_2+(1-x)Al → alloy Zr_xAl(1-x)+H_2↑. The samples were certified using: chemical analysis to determine the content of hydrogen (pyrolysis method); differential thermal analysis (DTA, derivatograph Q-1500, T_heating = 1000°C, rate 20°C/min); X-ray analysis (XRD, diffractometer DRON-0.5). The influences of the ratio of powders ZrH_2/Al in the reaction mixture, compacting pressure, temperature and heating velocity on the characteristics of the synthesized aluminides were determined. In HC, the solid solutions of Al in Zr, single phase ZrAl_2 and ZrAl_3 aluminides and Zr_3AlH_4.49 hydride were synthesized. Formation of aluminides in HC mode took place by the solid-phase mechanism, without melting of aluminum. During processing, the heating of the initial charge up to 540°C resulted in the decomposition of zirconium hydride (ZrH_2) to HCC ZrH_1.5, that interacted with aluminum at 630°C forming FCC alumohydride of zirconium. Further increase of the temperature up to 800°C led to complete decomposition of the formed alumohydride of zirconium. The final formation of the zirconium aluminide occurred at 1000-1100°C in the end of HC process. Conclusion: in the synthesis of zirconium aluminides, the HC mode has several significant advantages over the conventional modes: lower operating temperatures (1000°C instead of 1800°C); shorter duration (1.5-2 hours instead of tens of hours); the availability of

  17. Determination of hydrogen in zirconium hydride and uranium-zirconium hydride by inert gas exraction-gravimetric method

    International Nuclear Information System (INIS)

    Hoshino, Akira; Iso, Shuichi

    1976-01-01

    An inert gas extraction-gravimetric method has been applied to the determination of hydrogen in zirconium hydride and uranium-zirconium hydride which are used as neutron moderator and fuel of nuclear safety research reactor (NSRR), respectively. The sample in a graphite-enclosed quartz crucible is heated inductively to 1200 0 C for 20 min in a helium stream. Hydrogen liberated from the sample is oxidized to water by copper(I) oxide-copper(II) oxide at 400 0 C, and the water is determined gravimetrically by absorption in anhydrone. The extraction curves of hydrogen for zirconium hydride and uranium-zirconium hydride samples are shown in Figs. 2 and 3. Hydrogen in the samples is extracted quantitatively by heating at (1000 -- 1250) 0 C for (10 -- 40) min. Recoveries of hydrogen in the case of zirconium hydride were examined as follows: a weighed zirconium rod (5 phi x 6 mm, hydrogen -5 Torr. After the chamber was filled with purified hydrogen to 200 Torr, the rod was heated to 400 0 C for 15 h, and again weighed to determine the increase in weight. Hydrogen in the rod was then determined by the proposed method. The results are in excellent agreement with the increase in weight as shown in Table 1. Analytical results of hydrogen in zirconium hydride samples and an uranium-zirconium hydride sample are shown in Table 2. (auth.)

  18. Stress analysis of pressurized water reactor steam generator tube denting phenomena. Interim report

    International Nuclear Information System (INIS)

    Thomas, J.M.; Cipolla, R.C.; Ranjan, G.V.; Derbalian, G.

    1978-07-01

    In some Pressurized Water Reactor (PWR) steam generators, a corrosion product has formed on the carbon steel support plate in the crevice between the tube and support plate. The corrosion product occupies more volume than the original metal; the tube-to-support plate crevice volume is thus consumed with corrosion product, and further corrosive action results in a radially inward force on the tube and a radially outward force on the corroding support plate. This has resulted in indentation (''denting'') of the tube, accompanied by occasional cracking. Large in-plane deformation and cracking of support plates has also been observed in the most severely affected plants along with some serious side effects, such as deformation and cracking of inner row tube U-bends caused by support plate movement. Mechanical aspects of the tube denting phenomena have been studied using analytical models. The models used ranged from closed form analytical solutions to state-of-the-art numerical elastic-plastic computer program for moderate strains. It was found that tube dents, such as those observed in operating steam generators, are associated with yielding of both the tubes and support plates. Also studied were the stresses in tube U-bends caused by support plate flow slot deformation

  19. Optimization of drift gases for accuracy in pressurized drift tubes

    CERN Document Server

    Kirchner, J J; Dinner, A R; Fidkowski, K J; Wyatt, J H

    2001-01-01

    Modern detectors such as ATLAS use pressurized drift tubes to minimize diffusion and achieve high coordinate accuracy. However, the coordinate accuracy depends on the exact knowledge of converting measured times into coordinates. Linear space-time relationships are best for reconstruction, but difficult to achieve in the $E \\propto \\frac{1}{r}$ field. Previous mixtures, which contained methane or other organic quenchers, are disfavored because of ageing problems. From our studies of nitrogen and carbon dioxide, two mixtures with only small deviations from linearity were determined and measured. Scaling laws for different pressures and magnetic fields are also given.

  20. Optimization of drift gases for accuracy in pressurized drift tubes

    International Nuclear Information System (INIS)

    Kirchner, J.J.; Becker, U.J.; Dinner, R.B.; Fidkowski, K.J.; Wyatt, J.H.

    2001-01-01

    Modern detectors such as ATLAS use pressurized drift tubes to minimize diffusion and achieve high coordinate accuracy. However, the coordinate accuracy depends on the exact knowledge of converting measured times into coordinates. Linear space-time relationships are best for reconstruction, but difficult to achieve in the E∝1/r field. Previous mixtures, which contained methane or other organic quenchers, are disfavored because of ageing problems. From our studies of nitrogen and carbon dioxide, two mixtures with only small deviations from linearity were determined and measured. Scaling laws for different pressures and magnetic fields are also given

  1. Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor

    International Nuclear Information System (INIS)

    Hareux, F.; Roche, R.; Vrillon, B.

    1964-01-01

    Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [fr

  2. Material development for India's nuclear power programme

    Indian Academy of Sciences (India)

    rials with emphasis on development of fabrication routes of zirconium alloys for .... nuclear power programme, which envisages design and construction of thermal breeder ... Production of Hf-free nuclear grade zirconium ..... Later on for pressure tubes specified limit for hydrogen content in the as manufactured condition.

  3. Modification in band gap of zirconium complexes

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Mayank, E-mail: mayank30134@gmail.com; Singh, J.; Chouhan, S. [Department of Physics, ISLE, IPS Academy, Indore (M.P.) (India); Mishra, A. [School of Physics, Devi Ahilya Vishwavidyalaya, Indore (M.P.) (India); Shrivastava, B. D. [Govt. P. G. College, Biora (M.P.) (India)

    2016-05-06

    The optical properties of zirconium complexes with amino acid based Schiff bases are reported here. The zirconium complexes show interesting stereo chemical features, which are applicable in organometallic and organic synthesis as well as in catalysis. The band gaps of both Schiff bases and zirconium complexes were obtained by UV-Visible spectroscopy. It was found that the band gap of zirconium complexes has been modified after adding zirconium compound to the Schiff bases.

  4. Characteristics of CANDU fuel bundles that caused pressure tube fretting at the bundle midplane

    Energy Technology Data Exchange (ETDEWEB)

    Dennier, D; Manzer, A M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Koehn, E [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    Detailed measurements on new bundles, and those that caused fretting during in- and out-reactor tests, have given insight into the factors responsible for fretting at the midplane of the inlet bundle. Bottom fuel elements that were attached near radial endplate spokes and had inboard bearing pads in the rolled joint cavity produced a significant portion of the observed fret marks. These elements are influenced by several driving forces that deflect the centre bearing pads towards the pressure tube surface. The evidence suggests that slight changes in bundle design may be possible to reduce pressure tube fretting. (author). 4 refs., 3 tabs., 8 figs.

  5. Effects on Vocal Fold Collision and Phonation Threshold Pressure of Resonance Tube Phonation with Tube End in Water

    Science.gov (United States)

    Enflo, Laura; Sundberg, Johan; Romedahl, Camilla; McAllister, Anita

    2013-01-01

    Purpose: Resonance tube phonation in water (RTPW) or in air is a voice therapy method successfully used for treatment of several voice pathologies. Its effect on the voice has not been thoroughly studied. This investigation analyzes the effects of RTPW on collision and phonation threshold pressures (CTP and PTP), the lowest subglottal pressure…

  6. Plasma arc melting of zirconium

    International Nuclear Information System (INIS)

    Tubesing, P.K.; Korzekwa, D.R.; Dunn, P.S.

    1997-01-01

    Zirconium, like some other refractory metals, has an undesirable sensitivity to interstitials such as oxygen. Traditionally, zirconium is processed by electron beam melting to maintain minimum interstitial contamination. Electron beam melted zirconium, however, does not respond positively to mechanical processing due to its large grain size. The authors undertook a study to determine if plasma arc melting (PAM) technology could be utilized to maintain low interstitial concentrations and improve the response of zirconium to subsequent mechanical processing. The PAM process enabled them to control and maintain low interstitial levels of oxygen and carbon, produce a more favorable grain structure, and with supplementary off-gassing, improve the response to mechanical forming

  7. Laser-Doppler vibrating tube densimeter for measurements at high temperatures and pressures

    International Nuclear Information System (INIS)

    Aida, Tsutomu; Yamazaki, Ai; Akutsu, Makoto; Ono, Takumi; Kanno, Akihiro; Hoshina, Taka-aki; Ota, Masaki; Watanabe, Masaru; Sato, Yoshiyuki; Smith, Richard L. Jr.; Inomata, Hiroshi

    2007-01-01

    A laser-Doppler vibrometer was used to measure the vibration of a vibrating tube densimeter for measuring P-V-T data at high temperatures and pressures. The apparatus developed allowed the control of the residence time of the sample so that decomposition at high temperatures could be minimized. A function generator and piezoelectric crystal was used to excite the U-shaped tube in one of its normal modes of vibration. Densities of methanol-water mixtures are reported for at 673 K and 40 MPa with an uncertainty of 0.009 g/cm 3

  8. Low in reactor creep Zr-base alloy tubes

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Holt, R.A.

    1984-01-01

    This invention relates to zirconium alloy tubes especially for use in nuclear power reactors. More particularly it relates to quaternary 3.5 percent Sn, 1 percent Mo, 1 percent Nb, balance Zr alloy tubes which have been extruded, cold worked and heat treated to lower their dislocation density. In one embodiment the alloys are cold worked less than 5 percent and stress relieved to produce a low dislocation density and in another embodiment the alloys are cold worked up to about 50 percent and annealed to produce a very low dislocation density and also small equiaxed β grains

  9. Evaporation of R134a in a horizontal herringbone microfin tube: heat transfer and pressure drop

    Energy Technology Data Exchange (ETDEWEB)

    Wellsandt, S; Vamling, L [Chalmers University of Technology, Gothenburg (Sweden). Department of Chemical Engineering and Environmental Science, Heat and Power Technology

    2005-09-01

    An experimental investigation of in-tube evaporation of R134a has been carried out for a 4 m long herringbone microfin tube with an outer diameter of 9.53 mm. Measured local heat transfer coefficients and pressure losses are reported for evaporation temperatures between -0.7 and 10.1 {sup o}C and mass flow rates between 162 and 366 kg m{sup -2} s{sup -1}. Results from this work are compared to experimental results from literature as well as predicted values from some available helical microfin correlations. Differences in heat transfer mechanisms between helical and herringbone microfin tubes are discussed, as heat transfer coefficients in the investigated herringbone tube tend to peak at lower vapour qualities compared to helical microfins. Correlations developed for helical microfin tubes generally predict experimental values within {+-}30% for vapour qualities below 50%. However, at higher qualities none of the correlations are able to reflect the early peak of heat transfer coefficients. Predicted pressure gradients reproduce measured values in general within {+-}20%. (author)

  10. 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)

  11. Analytical tools and methodologies for evaluation of residual life of contacting pressure tubes in the early generation of Indian PHWRs

    International Nuclear Information System (INIS)

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

    2002-01-01

    In-service life of a contacting Zircaloy-2 pressure tube (PT) in the earlier generation of Indian PHWRs, is limited mainly due to the accelerated hydrogen pick-up and nucleation and growth of hydride blister(s) at the cold spot(s) formed on outside surface of pressure tube as a result of its contact with the calandria tube (CT). The activities involving development of the analytical models for simulating the degradation mechanisms leading to PT-CT contact and the methodologies for the revaluation of their safe life under such condition form the important part of our extensive programme for the life management of contacting pressure tubes. Since after the PT-CT contact, rate of hydrogen pick-up and nucleation and growth of hydride blisters govern the safe residual life of the pressure tube, two analytical models (a) hydrogen pick-up model ('HYCON') and (b) model for the nucleation and growth of hydride blister at the contact spot ('BLIST -2D') have been developed in-house to estimate the extent of degradation caused by them. Along with them, a methodology for evaluation of safe residual life has also been formulated for evaluating the safe residual life of the contacting channels. This paper gives the brief description of the models and the methodologies relevant for the contacting Zircaloy-2 pressure tubes. (author)

  12. Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

    1997-07-01

    Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

  13. Zirconium alloy barrier having improved corrosion resistance

    International Nuclear Information System (INIS)

    Adamson, R.B.; Rosenbaum, H.S.

    1983-01-01

    A nuclear fuel element for use in the core of a nuclear reactor has a composite cladding container having a substrate and a dilute zirconium alloy liner bonded to the inside surface of the substrate. The dilute zirconium alloy liner forms about 1 to about 20 percent of the thickness of the cladding and is comprised of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper. The dilute zirconium alloy liner shields the substrate from impurities or fission products from the nuclear fuel material and protects the substrate from stress corrosion and stress cracking. The dilute zirconium alloy liner displays greater corrosion resistance, especially to oxidation by hot water or steam than unalloyed zirconium. The substrate material is selected from conventional cladding materials, and preferably is a zirconium alloy. (author)

  14. Numerical study of pressure drop and heat transfer from circular and cam-shaped tube bank in cross-flow of nanofluid

    International Nuclear Information System (INIS)

    Mirabdolah Lavasani, Arash; Bayat, Hamidreza

    2016-01-01

    Highlights: • Flow around non-circular and circular shaped tube bank is studied. • Effect of using Al_2O_3-water nanofluid on flow and heat transfer is discussed. • Tubes are with in-line and staggered arrangement. • Pressure drop of non-circular tube is noticeably lower that circular tube. - Abstract: Flow and heat transfer of nanofluid inside circular and cam-shaped tube bank is studied numerically. Reynolds number for cam-shaped tube bank is defined based on equivalent diameter of circular tube and varies in range of 100 ⩽ Re_D ⩽ 400. Nanofluid is made by adding Al_2O_3 nanoparticle with volume fraction of 1–7% to pure water. Results show using nanofluid results in higher heat transfer rate for both circular tube bank and cam-shaped tube bank. Also, staggered arrangement has higher heat transfer for both circular and cam-shaped tube bank. Pressure drop from cam-shaped tube bank is substantially lower than circular tube bank for all range of Reynolds number and volume fraction.

  15. Proper Orthogonal Decomposition of Pressure Fields in a Draft Tube Cone of the Francis (Tokke) Turbine Model

    International Nuclear Information System (INIS)

    Stefan, D; Rudolf, P

    2015-01-01

    The simulations of high head Francis turbine model (Tokke) are performed for three operating conditions - Part Load, Best Efficiency Point (BEP) and Full Load using software Ansys Fluent R15 and alternatively OpenFOAM 2.2.2. For both solvers the simulations employ Realizable k-e turbulence model. The unsteady pressure pulsations of pressure signal from two monitoring points situated in the draft tube cone and one behind the guide vanes are evaluated for all three operating conditions in order to compare frequencies and amplitudes with the experimental results. The computed velocity fields are compared with the experimental ones using LDA measurements in two locations situated in the draft tube cone. The proper orthogonal decomposition (POD) is applied on a longitudinal slice through the draft tube cone. The unsteady static pressure fields are decomposed and a spatio-temporal behavior of modes is correlated with amplitude-frequency results obtained from the pressure signal in monitoring points. The main application of POD is to describe which modes are related to an interaction between rotor (turbine runner) and stator (spiral casing and guide vanes) and cause dynamic flow behavior in the draft tube. The numerically computed efficiency is correlated with the experimental one in order to verify the simulation accuracy

  16. Study the Effect of Nasal Obstruction Surgery (Septoplasty on Eustachian Tube Function and Middle Ear Pressure

    Directory of Open Access Journals (Sweden)

    Rahim Davari

    2014-12-01

    Full Text Available Background & objectives: Deviation of the nasal septum is a common cause of unilateral or bilateral nasal airway obstruction and may follow nasal and midface trauma. Patient complaints of airway obstruction that are consistent with intranasal physical findings often lead to septoplasty and turbinate surgery. Severe nasal septal deviation leads to complete nasal obstruction and disturbs air passage from nostrils, however the effect of septal deviation and nasal obstruction surgery (septoplasty on Eustachian tube function and middle ear pressure is controversial and isn’t clear. Whereas many of surgeons do not believe the considerable effect of septoplasty on Eustachian tube function and middle ear pressure, so they do not recommend this procedure before middle ear surgery. On the other hand, some have an idea that septoplasty has significant effect on middle ear pressure suggesting this procedure before ear surgery like tympanoplasty.   Methods: This prospective analytical-descriptive study was conducted on seventy patients from 18 to 65 years of age who underwent septoplasty due to severe septal deviation leading to nasal obstruction in Beesat Hospital during one year (2012-2013. Middle ear pressure and Eustachian tube function on the septal deviated side and contralateral side before and after septoplasty (3 to 6 months later were measured through tympanometry and Eustachian tube function test (Toynbee test. The comparison between pre- and postoperative ETF tests and middle ear pressures was assessed using Paired –T test and p-value of less than 0.05 was considered as statistically significant.   Results : This study revealed that comparison between mean values of pressure in the deviated and contralateral side has no significant statistical difference before and after septoplasty. Also comparison between Eustachian function in the deviated side and contralateral side showed no significant difference before and after septoplasty.

  17. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1984-11-01

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

  18. Some observations of the pressure distribution in a tube bank for conditions of self generated acoustic resonance

    International Nuclear Information System (INIS)

    Fitzpatrick, J.A.; Donaldson, I.S.; McKnight, W.

    1979-01-01

    The results for mean and fluctuating pressure distributions around tubes in an in-line tube bank are presented for both non-resonant and self-excited acoustic standing wave resonant flow regimes. It is readily deduced that the nature of the flow in the bank is dramatically altered with the onset of acoustic resonance. The velocity gradients which appear across the bank with the onset of resonance would suggest regions of flow recirculation in the bank although no evidence of this was found. The spectra of fluctuating pressure on the duct roof in the bank and on tubes deep in the bank exhibited coherent peaks only during resonance. (author)

  19. Joining method for pressure tube and martensitic stainless steel tube

    International Nuclear Information System (INIS)

    Kimoto, Hiroshi; Koike, Hiromitsu.

    1993-01-01

    In a joining portion of zirconium alloy and a stainless steel, the surface of martensitic stainless steel being in contact with Zr and Zr alloy is applied with a laser quenching solidification treatment before expanding joining of them to improve the surface. This can provide the surface with refined coagulated cell tissues and make deposits and impurities homogeneous and solubilized. As a result, the surface of the martensitic stainless steel has highly corrosion resistance, to suppress contact corrosion with Zr and Zr alloy. Accordingly, even if it is exposed to high temperature water of 200 to 350degC, failures of Zr and Zr alloy can be suppressed. (T.M.)

  20. Boiling heat transfer and dryout in helically coiled tubes under different pressure conditions

    International Nuclear Information System (INIS)

    Chung, Young-Jong; Bae, Kyoo-Hwan; Kim, Keung Koo; Lee, Won-Jae

    2014-01-01

    Highlights: • Heat transfer characteristics and dryout for helically coiled tube are performed. • A boiling heat transfer tends to increase with a pressure increase. • Dryout occurs at high quality test conditions investigated. • Steiner–Taborek’s correlation is predicted well based on the experimental results. - Abstract: A helically coiled once-through steam generator has been used widely during the past several decades for small nuclear power reactors. The heat transfer characteristics and dryout conditions are important to optimal design a helically coiled steam generator. Various experiments with the helically coiled tubes are performed to investigate the heat transfer characteristics and occurrence condition of a dryout. For the investigated experimental conditions, Steiner–Taborek’s correlation is predicted reasonably based on the experimental results. The pressure effect is important for the boiling heat transfer correlation. A boiling heat transfer tends to increase with a pressure increase. However, it is not affected by the pressure change at a low power and low mass flow rate. Dryout occurs at high quality test conditions investigated because a liquid film on the wall exists owing to a centrifugal force of the helical coil

  1. Zirconium behaviour in purex process solutions

    International Nuclear Information System (INIS)

    Shu, J.

    1982-01-01

    The extraction behaviour of zirconium, as fission product, in TBP/diluent- HNO 3 -H 2 O systems, simulating Purex solutions, is studied. The main purpose is to attain an increasing in the zirconium decontamination factor by adjusting the extraction parameters. Equilibrium diagram, TBP concentration, aqueous:organic ratio, salting-out effects and, uranium loading in the organic phase were the main factors studied. All these experiments had been made with zirconium in the 10 - 2 - 10 - 3 concentration range. The extractant degradation products influence uppon the zirconium behaviour was also verified. With the obtained data it was possible to introduce some modification in the standard Purex flow-sheet in order to obtain the uranium product with higher zirconium decontamination. (Author) [pt

  2. Pre and post garter spring repositioning ultrasonic inspection of pressure tubes

    International Nuclear Information System (INIS)

    Desimone, C.; Katchadjian, P.; Tacchia, Mauricio

    1997-01-01

    This paper present a description of the ultrasonic cracked hydride blister detections system used for pre and post inspection of pressure tubes during garter spring repositioning in CNE (Embalse Nuclear Power Station). Ultrasonic system setup configuration, transducers characteristics, blister detection head, calibration of parameters, operating procedure, records of ultrasonic inspections and evaluation. (author) [es

  3. Condensation heat transfer and pressure drop of R-410A in flat aluminum multi-port tubes

    Science.gov (United States)

    Kim, Nae-Hyun

    2018-02-01

    Brazed heat exchangers with aluminum flat multi-port tubes are being used as condensers of residential air-conditioners. In this study, R-410A condensation tests were conducted in four multi-port tubes having a range of hydraulic diameter (0.78 ≤ Dh ≤ 0.95 mm). The test range covered the mass flux from 100 to 400 kg/m2 s and the heat flux at 3 kW/m2, which are typical operating conditions of residential air conditioners. Results showed that both the heat transfer coefficient and the pressure drop increased as the hydraulic diameter decreased. The effect of hydraulic diameter on condensation heat transfer was much larger than the predictions of existing correlations for the range of investigation. Comparison of the data with the correlations showed that some macro-channel tube correlations and mini-channel tube correlations reasonably predicted the heat transfer coefficient. However, macro-channel correlations highly overpredicted the pressure drop data.

  4. Miniaturised Prandtl tube with integrated pressure sensors for micro-thruster plume characterisation

    NARCIS (Netherlands)

    Dijkstra, Marcel; Ma, Kechun; de Boer, Meint J.; Groenesteijn, Jarno; Lötters, Joost Conrad; Wiegerink, Remco J.

    2014-01-01

    A miniaturised Prandtl-tube sensor incorporating a 6 mm long 40 μm diameter microchannel with integrated pressure sensors has been realised. The sensor has been designed for the characterisation of rarefied plume flow from a MEMS-based monopropellant propulsion system for high-accuracy attitude

  5. 2D modeling of moderator flow and temperature distribution around a single channel after pressure tube/calandria tube contact

    International Nuclear Information System (INIS)

    Behdadi, A.; Luxat, J.C.

    2009-01-01

    A 2D computational fluid dynamics (CFD) model has been developed to calculate the moderator velocity field and temperature distribution around a single channel inside the moderator of a CANDU reactor after a postulated ballooning deformation of the pressure tube (PT) into contact with the calandria tube (CT). Following contact between the hot PT and the relatively cold CT, there is a spike in heat flux to the moderator surrounding the CT which may lead to sustained CT dryout. This can detrimentally affect channel integrity if the CT post-dryout temperature becomes sufficiently high to result in thermal creep strain deformation. The present research is focused on establishing the limits for dryout occurrence on the CTs for the situation in which pressure tube-calandria tube contact occurs. In order to consider different location of the channels inside the calandria, both upward and downward flow directions have been analyzed. The standard κ - ε turbulence model associated with logarithmic wall function is applied to predict the effects of turbulence. The governing equations are solved by the finite element software package COMSOL. The buoyancy driven natural convection on the outer surface of a CT has been analyzed to predict the flow and temperature distribution around the single CT considering the local moderator subcooling, wall temperature and heat flux. The model also shows the effect of high CT temperature on the flow and subcooling around the CTs at higher/lower elevation depending on the flow direction in the domain. According to the flow pattern and temperature distribution, it is predicted that stable film boiling generates in the stagnation region on the cylinder. (author)

  6. The study on pressure oscillation and heat transfer characteristics of oscillating capillary tube heat pipe

    International Nuclear Information System (INIS)

    Kim, Jong Soo; Bui, Ngoc Hung; Jung, Hyun Seok; Lee, Wook Hyun

    2003-01-01

    In the present study, the characteristics of pressure oscillation and heat transfer performance in an oscillating capillary tube heat pipe were experimentally investigated with respect to the heat flux, the charging ratio of working fluid, and the inclination angle to the horizontal orientation. The experimental results showed that the frequency of pressure oscillation was between 0.1 Hz and 1.5 Hz at the charging ratio of 40 vol.%. The saturation pressure of working fluid in the oscillating capillary tube heat pipe increased as the heat flux was increased. Also, as the charging ratio of working fluid was increased, the amplitude of pressure oscillation increased. When the pressure waves were symmetric sinusoidal waves at the charging ratios of 40 vol.% and 60 vol.%, the heat transfer performance was improved. At the charging ratios of 20 vol.% and 80 vol.%, the waveforms of pressure oscillation were more complicated, and the heat transfer performance reduced. At the charging ratio of 40 vol.%, the heat transfer performance of the OCHP was at the best when the inclination angle was 90 .deg., the pressure wave was a sinusoidal waveform, the pressure difference was at the least, the oscillation amplitude was at the least, and the frequency of pressure oscillation was the highest

  7. Critical heat flux of water in vertical round tubes at low-pressure and low-flow conditions

    International Nuclear Information System (INIS)

    Park, Jae-Wook; Kim, Hong-Chae; Beak, Won-Pil; Chang, Soon Heung

    1997-01-01

    A series of critical heat flux (CHF) tests have been performed to provide a reliable set of CHF data for water flow in vertical round tubes at low pressure and low flow (LPLF) conditions. The range of experimental conditions is as follows: diameter 8, 10 mm; heated length 0.5, 1 m; pressure 2-9 bar, mass flux 50-200 kg/m 2 s; inlet subcooling 350, 450 kJ/kg. The observed parametric trends are generally consistent with the previous understanding except for the effects of system pressure and tube diameter. The pressure effect is small but very complicated; existing CHF correlations do not represent this parametric trend properly. CHF increases with the increase in diameter at fixed exit conditions, contrary to the general understanding. The artificial neural networks are applied to the round tube CHF data base at LPLF (P = 110-1100 kPa, G = 0-500 kg/m 2 s) conditions. The trained backpropagation networks (BPNs) predict CHF better than any other CHF correlations. Parametric trends of CHF based on the BPN for fixed inlet conditions generally agree well with our experimental results. (author)

  8. A novel method for calculating the dynamic capillary force and correcting the pressure error in micro-tube experiment.

    Science.gov (United States)

    Wang, Shuoliang; Liu, Pengcheng; Zhao, Hui; Zhang, Yuan

    2017-11-29

    Micro-tube experiment has been implemented to understand the mechanisms of governing microcosmic fluid percolation and is extensively used in both fields of micro electromechanical engineering and petroleum engineering. The measured pressure difference across the microtube is not equal to the actual pressure difference across the microtube. Taking into account the additional pressure losses between the outlet of the micro tube and the outlet of the entire setup, we propose a new method for predicting the dynamic capillary pressure using the Level-set method. We first demonstrate it is a reliable method for describing microscopic flow by comparing the micro-model flow-test results against the predicted results using the Level-set method. In the proposed approach, Level-set method is applied to predict the pressure distribution along the microtube when the fluids flow along the microtube at a given flow rate; the microtube used in the calculation has the same size as the one used in the experiment. From the simulation results, the pressure difference across a curved interface (i.e., dynamic capillary pressure) can be directly obtained. We also show that dynamic capillary force should be properly evaluated in the micro-tube experiment in order to obtain the actual pressure difference across the microtube.

  9. Anisotropy of mechanical properties of zirconium and zirconium alloys

    International Nuclear Information System (INIS)

    Medrano, R.E.

    1975-01-01

    In studies of technological applications of zirconium to fuel elements of nuclear reactor, it was found that the use of plasticity equations for isotropic materials is not in agreement with experimental results, because of the strong anisotropy of zirconium. The present review describes recent progress on the knowledge of the influence of anisotropy on mechanical properties, after Douglass' review in 1971. The review was written to be selfconsistent, changing drastically the presentation of some of the referenced papers. It is also suggested some particular experiments to improve developments in this area

  10. Investigation of forced convection heat transfer of supercritical pressure water in a vertically upward internally ribbed tube

    International Nuclear Information System (INIS)

    Wang Jianguo; Li Huixiong; Guo Bin; Yu Shuiqing; Zhang Yuqian; Chen Tingkuan

    2009-01-01

    In the present paper, the forced convection heat transfer characteristics of water in a vertically upward internally ribbed tube at supercritical pressures were investigated experimentally. The six-head internally ribbed tube is made of SA-213T12 steel with an outer diameter of 31.8 mm and a wall thickness of 6 mm and the mean inside diameter of the tube is measured to be 17.6 mm. The experimental parameters were as follows. The pressure at the inlet of the test section varied from 25.0 to 29.0 MPa, and the mass flux was from 800 to 1200 kg/(m 2 s), and the inside wall heat flux ranged from 260 to 660 kW/m 2 . According to experimental data, the effects of heat flux and pressure on heat transfer of supercritical pressure water in the vertically upward internally ribbed tube were analyzed, and the characteristics and mechanisms of heat transfer enhancement, and also that of heat transfer deterioration, were also discussed in the so-called large specific heat region. The drastic changes in thermophysical properties near the pseudocritical points, especially the sudden rise in the specific heat of water at supercritical pressures, may result in the occurrence of the heat transfer enhancement, while the covering of the heat transfer surface by fluids lighter and hotter than the bulk fluid makes the heat transfer deteriorated eventually and explains how this lighter fluid layer forms. It was found that the heat transfer characteristics of water at supercritical pressures were greatly different from the single-phase convection heat transfer at subcritical pressures. There are three heat transfer modes of water at supercritical pressures: (1) normal heat transfer, (2) deteriorated heat transfer with low HTC but high wall temperatures in comparison to the normal heat transfer, and (3) enhanced heat transfer with high HTC and low wall temperatures in comparison to the normal heat transfer. It was also found that the heat transfer deterioration at supercritical pressures was

  11. Turbulent convective heat transfer of methane at supercritical pressure in a helical coiled tube

    Science.gov (United States)

    Wang, Chenggang; Sun, Baokun; Lin, Wei; He, Fan; You, Yingqiang; Yu, Jiuyang

    2018-02-01

    The heat transfer of methane at supercritical pressure in a helically coiled tube was numerically investigated using the Reynolds Stress Model under constant wall temperature. The effects of mass flux ( G), inlet pressure ( P in) and buoyancy force on the heat transfer behaviors were discussed in detail. Results show that the light fluid with higher temperature appears near the inner wall of the helically coiled tube. When the bulk temperature is less than or approach to the pseudocritical temperature ( T pc ), the combined effects of buoyancy force and centrifugal force make heavy fluid with lower temperature appear near the outer-right of the helically coiled tube. Beyond the T pc , the heavy fluid with lower temperature moves from the outer-right region to the outer region owing to the centrifugal force. The buoyancy force caused by density variation, which can be characterized by Gr/ Re 2 and Gr/ Re 2.7, enhances the heat transfer coefficient ( h) when the bulk temperature is less than or near the T pc , and the h experiences oscillation due to the buoyancy force. The oscillation is reduced progressively with the increase of G. Moreover, h reaches its peak value near the T pc . Higher G could improve the heat transfer performance in the whole temperature range. The peak value of h depends on P in. A new correlation was proposed for methane at supercritical pressure convective heat transfer in the helical tube, which shows a good agreement with the present simulated results.

  12. Numerical studies on heat transfer and pressure drop characteristics of flat finned tube bundles with various fin materials

    Science.gov (United States)

    Peng, Y.; Zhang, S. J.; Shen, F.; Wang, X. B.; Yang, X. R.; Yang, L. J.

    2017-11-01

    The air-cooled heat exchanger plays an important role in the field of industry like for example in thermal power plants. On the other hand, it can be used to remove core decay heat out of containment passively in case of a severe accident circumstance. Thus, research on the performance of fins in air-cooled heat exchangers can benefit the optimal design and operation of cooling systems in nuclear power plants. In this study, a CFD (Computational Fluid Dynamic) method is implemented to investigate the effects of inlet velocity, fin spacing and tube pitch on the flow and the heat transfer characteristics of flat fins constructed of various materials (316L stainless steel, copper-nickel alloy and aluminium). A three dimensional geometric model of flat finned tube bundles with fixed longitudinal tube pitch and transverse tube pitch is established. Results for the variation of the average convective heat transfer coefficient with respect to cooling air inlet velocity, fin spacing, tube pitch and fin material are obtained, as well as for the pressure drop of the cooling air passing through finned tube. It is shown that the increase of cooling air inlet velocity results in enhanced average convective heat transfer coefficient and decreasing pressure drop. Both fin spacing and tube pitch engender positive effects on pressure drop and have negative effects on heat transfer characteristics. Concerning the fin material, the heat transfer performance of copper-nickel alloy is superior to 316L stainless steel and inferior to aluminium.

  13. Denting of Inconel 600 steam generator tubes in pressurized water reactors

    International Nuclear Information System (INIS)

    Van Rooyen, D.; Weeks, J.R.

    1976-10-01

    Rapid, localized corrosion of carbon steel tube support plates (TSP) has led to cases of denting of steam generator tubes, due to the pressure of corrosion products formed in crevices between the tubes and TSP holes. The corrosion product is mainly magnetite (Fe 3 O 4 ), formed in ''run-away'' fashion as a result of local chemistry changes when an extended operation with phosphate (PO 4 ) treatment of the secondary coolant is followed by an all volatile treatment (AVT). The rate of the ''run-away'' magnetite formation, and therefore, the extent of damage will probably vary with the amounts of the harmful chemicals present and with temperature. Leaky condensers are felt to be responsible for the presence of Cl - ions, and for the observation that denting is more extensive in plants with salt water cooled condensers. It is possible that thermal cycles assist the denting process, both by mechanical and chemical ratchetting mechanisms. Recommendations are presented concerning the continued operation of plants with observed denting

  14. 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)

  15. Photoelectrochemical properties and band structure of oxide films on zirconium-transition metal alloys

    International Nuclear Information System (INIS)

    Takahashi, Kazuo; Uno, Masayoshi; Okui, Mihoko; Yamanaka, Shinsuke

    2006-01-01

    The microalloying effects of 4d and 5d transition metals, M (M: Nb, Mo, Ta, W) on the photoelectrochemical properties, the flat band potential (U fb ) and the band gap energy (E g ), for zirconium oxide films were investigated by photoelectrochemical measurements and band calculation. Button ingots of zirconium-5 mol% M (M: Nb, Mo, Ta, W) were made from high-purity metals (99.9% purity) by arc melting in a purified argon atmosphere. These plate specimens were sealed into silica tubes in vacuum, and then homogenized at 1273 K for 24 h. Subsequently, these specimens were oxidized up to 1173 K. The photocurrent of each specimen was evaluated at room temperature under the irradiation of Xe lamp (500 W) through grating monochrometer and cut-off filter. 0.1 M Na 2 SO 4 solution was used as the electrolyte. The value of the flat band potential was higher and the value of the band gap energy was smaller than that of pure zirconium oxide film in all sample. It was found from the calculation by CASTEP code that the decreases in band gap energy of these oxide films was due to formation of 4d or 5d orbital of transition metals

  16. Reconstruction of an acoustic pressure field in a resonance tube by particle image velocimetry.

    Science.gov (United States)

    Kuzuu, K; Hasegawa, S

    2015-11-01

    A technique for estimating an acoustic field in a resonance tube is suggested. The estimation of an acoustic field in a resonance tube is important for the development of the thermoacoustic engine, and can be conducted employing two sensors to measure pressure. While this measurement technique is known as the two-sensor method, care needs to be taken with the location of pressure sensors when conducting pressure measurements. In the present study, particle image velocimetry (PIV) is employed instead of a pressure measurement by a sensor, and two-dimensional velocity vector images are extracted as sequential data from only a one- time recording made by a video camera of PIV. The spatial velocity amplitude is obtained from those images, and a pressure distribution is calculated from velocity amplitudes at two points by extending the equations derived for the two-sensor method. By means of this method, problems relating to the locations and calibrations of multiple pressure sensors are avoided. Furthermore, to verify the accuracy of the present method, the experiments are conducted employing the conventional two-sensor method and laser Doppler velocimetry (LDV). Then, results by the proposed method are compared with those obtained with the two-sensor method and LDV.

  17. Influence de l'orientation des hydrures sur les modes de déformation, d'endommagement et de rupture du Zircaloy-4 hydruré.

    OpenAIRE

    Racine , Aude

    2005-01-01

    In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300°C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embitterment depends on many parameters, among which hydrogen content and orientation of hydrid...

  18. Study on drop pressure and flow distribution of double-tube heat exchanger

    International Nuclear Information System (INIS)

    Liu Junqiang; Chen Minghui; Hu Yumin; Li Rizhu; Kong Dechun; Zhang Weijie

    2007-01-01

    The parallel connection channel pressure drop characters of the double-tube bundle heat exchange were experimentally investigated in this paper in order to find out how the flow of the heat exchanger is distributed and then to optimize the structure of heat exchanger according to the flow distribution. A double-tube bundle heat exchanger was built according to the similarity criteria. The experiment system was also built to test the optimization of the heat exchanger. The experiment results reveal that the calculating model is reliable and decreasing pipe space to optimize the heat exchanger is reasonable. (authors)

  19. New solvent extraction process for zirconium and hafnium

    International Nuclear Information System (INIS)

    Takahashi, M.; Katoh, Y.; Miyazaki, H.

    1984-01-01

    The authors' company developed a new solvent extraction process for zirconium and hafnium separation, and started production of zirconium sponge by this new process in September 1979. The process utilizes selective extraction of zirconium oxysulfate using high-molecular alkyl amine, and has the following advantages: 1. This extraction system has a separation factor as high as 10 to 20 for zirconium and hafnium in the range of suitable acid concentration. 2. In the scrubbing section, removal of all the hafnium that coexists with zirconium in the organic solvent can be effectively accomplished by using scrubbing solution containing hafnium-free zirconium sulfate. Consequently, hafnium in the zirconium sponge obtained is reduced to less than 50 ppm. 3. The extractant undergoes no chemical changes but is very stable for a long period. In particular, its solubility in water is small, about 20 ppm maximum, posing no environmental pollution problems such as are often caused by other process raffinates. At the present time, the zirconium and hafnium separation operation is very stable, and zirconium sponge made by this process can be applied satisfactorily to nuclear reactors

  20. Phase transitions in solids under high pressure

    CERN Document Server

    Blank, Vladimir Davydovich

    2013-01-01

    Phase equilibria and kinetics of phase transformations under high pressureEquipment and methods for the study of phase transformations in solids at high pressuresPhase transformations of carbon and boron nitride at high pressure and deformation under pressurePhase transitions in Si and Ge at high pressure and deformation under pressurePolymorphic α-ω transformation in titanium, zirconium and zirconium-titanium alloys Phase transformations in iron and its alloys at high pressure Phase transformations in gallium and ceriumOn the possible polymorphic transformations in transition metals under pressurePressure-induced polymorphic transformations in АIBVII compoundsPhase transformations in AIIBVI and AIIIBV semiconductor compoundsEffect of pressure on the kinetics of phase transformations in iron alloysTransformations during deformation at high pressure Effects due to phase transformations at high pressureKinetics and hysteresis in high-temperature polymorphic transformations under pressureHysteresis and kineti...

  1. Fine-grained zirconium-base material

    Science.gov (United States)

    Van Houten, G.R.

    1974-01-01

    A method is described for making zirconium with inhibited grain growth characteristics, by the process of vacuum melting the zirconium, adding 0.3 to 0.5% carbon, stirring, homogenizing, and cooling. (Official Gazette)

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

    International Nuclear Information System (INIS)

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

    1984-05-01

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

  3. Microhardness and microplasticity of zirconium nitride

    International Nuclear Information System (INIS)

    Neshpor, V.S.; Eron'yan, M.A.; Petrov, A.N.; Kravchik, A.E.

    1978-01-01

    To experimentally check the concentration dependence of microhardness of 4 group nitrides, microhardness of zirconium nitride compact samples was measured. The samples were obtained either by bulk saturation of zirconium iodide plates or by chemical precipitation from gas. As nitrogen content decreased within the limits of homogeneity of zirconium nitride samples where the concentration of admixed oxygen was low, the microhardness grew from 1500+-100 kg/mm 2 for ZrNsub(1.0) to 27000+-100 kg/mm 2 for ZrNsub(0.78). Microplasticity of zirconium nitride (resistance to fracture) decreased, as the concentration of nitrogen vacancies was growing

  4. Manufacture of seamless stainless steel tubings and related equipment

    International Nuclear Information System (INIS)

    Wali, D.K.; Chaudhary, S.

    1997-01-01

    Production of seamless tubes for special application is one of the important production activities of Nuclear Fuel Complex (NFC), Hyderabad. NFC had set up facility of Hot Extrusion Press and Cold Pilger Mills with related finishing and inspection equipment for manufacturing quality seamless tubes of zirconium alloy for application in nuclear power reactors in early 70''s. Being aware that the demand for seamless tube in a developing economy gradually increases till it reaches around 30 to 40% of the total requirement of tubes and pipes and also of the fact that manufacturing technology developed for production of zircaloy seamless tubes for nuclear application, can easily be harnessed and spinned off for production of seamless tubes in materials generally difficult to hot roll (in other than extrusion process), NFC augmented its seamless tube manufacturing facility by adding, a vertical piercing press, series of induction furnaces and large size pilger mills to meet existing market demand of power sector, engineering, fertilisers and petro chemical industries and any other specialised applications

  5. Characteristics of two-phase flow pattern transitions and pressure drop of five refrigerants in horizontal circular small tubes

    Energy Technology Data Exchange (ETDEWEB)

    Pamitran, A.S. [Department of Mechanical Engineering, University of Indonesia, Kampus Baru UI, Depok 16424 (Indonesia); Choi, Kwang-Il [Graduate School, Chonnam National University, San 96-1, Dunduk-Dong, Yeosu, Chonnam 550-749 (Korea); Oh, Jong-Taek [Department of Refrigeration and Air Conditioning Engineering, Chonnam National University, San 96-1, Dunduk-Dong, Yeosu, Chonnam 550-749 (Korea); Hrnjak, Pega [Department of Mechanical Science and Engineering, ACRC, University of Illinois at Urbana-Champaign, 1206 West Green Street, Urbana, IL 61801 (United States)

    2010-05-15

    An experimental investigation on the characteristics of two-phase flow pattern transitions and pressure drop of R-22, R-134a, R-410A, R-290 and R-744 in horizontal small stainless steel tubes of 0.5, 1.5 and 3.0 mm inner diameters is presented. Experimental data were obtained over a heat flux range of 5-40 kW/m{sup 2}, mass flux range of 50-600 kg/(m{sup 2} s), saturation temperature range of 0-15 C, and quality up to 1.0. Experimental data were evaluated with Wang et al. and Wojtan et al. [Wang, C.C., Chiang, C.S., Lu, D.C., 1997. Visual observation of two-phase flow pattern of R-22, R-134a, and R-407C in a 6.5-mm smooth tube. Exp. Therm. Fluid Sci. 15, 395-405; Wojtan, L., Ursenbacher, T., Thome, J.R., 2005. Investigation of flow boiling in horizontal tubes: part I - a new diabatic two-phase flow pattern map. Int. J. Heat Mass Transfer 48, 2955-2969.] flow pattern maps. The effects of mass flux, heat flux, saturation temperature and inner tube diameter on the pressure drop of the working refrigerants are reported. The experimental pressure drop was compared with the predictions from some existing correlations. A new two-phase pressure drop model that is based on a superposition model for two-phase flow boiling of refrigerants in small tubes is presented. (author)

  6. SEPARATION OF HAFNIUM FROM ZIRCONIUM

    Science.gov (United States)

    Overholser, L.B.; Barton, C.J. Sr.; Ramsey, J.W.

    1960-05-31

    The separation of hafnium impurities from zirconium can be accomplished by means of organic solvent extraction. The hafnium-containing zirconium feed material is dissolved in an aqueous chloride solution and the resulting solution is contacted with an organic hexone phase, with at least one of the phases containing thiocyanate. The hafnium is extracted into the organic phase while zirconium remains in the aqueous phase. Further recovery of zirconium is effected by stripping the onganic phase with a hydrochloric acid solution and commingling the resulting strip solution with the aqueous feed solution. Hexone is recovered and recycled by means of scrubbing the onganic phase with a sulfuric acid solution to remove the hafnium, and thiocyanate is recovered and recycled by means of neutralizing the effluent streams to obtain ammonium thiocyanate.

  7. Zirconium

    Science.gov (United States)

    Bedinger, G.M.

    2013-01-01

    Zirconium is the 20th most abundant element in the Earth’s crust. It occurs in a variety of rock types and geologic environments but most often in igneous rocks in the form of zircon (ZrSiO4). Zircon is recovered as a coproduct of the mining and processing of heavy mineral sands for the titanium minerals ilmenite and rutile. The sands are formed by the weathering and erosion of rock containing zircon and titanium heavy minerals and their subsequent concentration in sedimentary systems, particularly in coastal environments. A small quantity of zirconium, less than 10 kt/a (11,000 stpy), compared with total world production of 1.4 Mt (1.5 million st) in 2012, was derived from the mineral baddeleyite (ZrO2), produced from a single source in Kovdor, Russia.

  8. Experiments of draining and filling processes in a collapsible tube at high external pressure

    Science.gov (United States)

    Flaud, P.; Guesdon, P.; Fullana, J.-M.

    2012-02-01

    The venous circulation in the lower limb is mainly controlled by the muscular action of the calf. To study the mechanisms governing the venous draining and filling process in such a situation, an experimental setup, composed by a collapsible tube under external pressure, has been built. A valve preventing back flows is inserted at the bottom of the tube and allows to model two different configurations: physiological when the fluid flow is uni-directional and pathological when the fluid flows in both directions. Pressure and flow rate measurements are carried out at the inlet and outlet of the tube and an original optical device with three cameras is proposed to measure the instantaneous cross-sectional area. The experimental results (draining and filling with physiological or pathological valves) are confronted to a simple one-dimensional numerical model which completes the physical interpretation. One major observation is that the muscular contraction induces a fast emptying phase followed by a slow one controlled by viscous effects, and that a defect of the valve decreases, as expected, the ejected volume.

  9. Extractive metallurgy of zirconium--1945 to the present

    International Nuclear Information System (INIS)

    Franklin, D.G.; Adamson, R.B.

    1984-01-01

    Although the history of the reduction of zirconium dates from 1824 and the first ductile zirconium metal was produced in the laboratory in 1914, modern reduction practice was pioneered by the U.S. Bureau of Mines starting in 1945. This paper reviews the history of the extractive metallurgy of zirconium from the early work of W. J. Kroll and co-workers at the Bureau of Mines in Albany, Ore., through the commercial development of the production of reactor-grade zirconium metal which was spurred by the requirements of the Naval Reactor Program and the development of commercial nuclear power. Technical subjects covered include processes for opening the ore, zirconium-hafnium separation, chlorination of zirconium oxide, reduction processes, and electrowinning of zirconium metal. Proposed new processes and process modifications are reviewed

  10. Management of large scale coolant channel replacement programme for Indian PHWRs

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  11. Method for shaping polyethylene tubing

    Science.gov (United States)

    Kramer, R. C.

    1981-01-01

    Method forms polyethylene plastic tubing into configurations previously only possible with metal tubing. By using polyethylene in place of copper or stain less steel tubing inlow pressure systems, fabrication costs are significantly reduced. Polyethylene tubing can be used whenever low pressure tubing is needed in oil operations, aircraft and space applications, powerplants, and testing laboratories.

  12. Research into zirconium alloys resistant to carbon dioxide under pressure at temperatures of up to 600 deg C (1963); Recherche d'alliages de zirconium compatibles avec le gaz carbonique sous pression jusqu'a 500 ou 600 deg C (1063)

    Energy Technology Data Exchange (ETDEWEB)

    Baque, P; Dominget, R; Bossard, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    Zirconium is a metal having a relatively low neutron capture cross-section and a high melting point; it is thus possible to consider its use in particular as a canning material for fuel elements in CO{sub 2}-cooled nuclear reactors. A preliminary study of several types of zirconium showed that the metal is already strongly oxidised in this gas at 500 deg C. The 'breakaway' phenomenon is generalised; the oxidation rate is then linear and depends on the carbon dioxide pressure. An attempt was therefore made to find binary and tertiary alloys in order to improve the metal behaviour. Several interesting compositions were found: 1, 1.6 and 2.5 per cent of copper, 2 per cent of vanadium, and 0.05 and 0.5 per cent of calcium. Tertiary copper-molybdenum and copper-phosphorus alloys are also less liable to oxidation and in particular do not exhibit the 'breakaway' phenomenon even after a prolonged treatment at 600 deg C. (authors) [French] Le zirconium se trouve parmi les metaux a section de capture neutronique relativement faible et possede une temperature de fusion elevee; aussi peut on songer a l'employer notamment comme materiau de gainage d'elements combustibles pour reacteurs nucleaires refroidis au gaz carbonique. Une etude prealable de plusieurs qualites de zirconium a montre que le metal est deja assez fortement oxyde dans ce gaz des 500 deg C. En effet, le phenomene de ''breakaway'' est general; la vitesse d'oxydation devient alors lineaire et depend de la pression du gaz carbonique. La recherche d'alliages binaires et ternaires a donc ete entreprise afin de tenter d'ameliorer le comportement du metal. Elle a permis d'aboutir a quelques compositions interessantes: cuivre 1, 1,6 et 2,5 pour cent, vanadium 2 pour cent, et calcium 0,05 et 0,5 pour cent. Des alliages ternaires au cuivre-molybdene et cuivre-phosphore sont egalement moins oxydables, et en particulier ne presentent pas le phenomene de ''breakaway'', meme apres une longue exposition a 600 deg C. (auteurs)

  13. The degradation of zirconium alloys in nuclear reactors - a review

    International Nuclear Information System (INIS)

    Lim, D.; Graham, N.A.

    1986-01-01

    This report presents the findings of a survey of available non-Canadian literature on the oxidation and hydriding of zirconium alloys. Much of the literature was found to address the Zircaloys, particularly when used as fuel cladding subjected to a radioactive and oxidizing environment. Hydriding of Zircaloys is mainly attributed to oxidation. The survey revealed that Zr-Nb alloys have been included in some investigations; however, data on the long-term degradation of Zr-2.5 wt% Nb, in particular, were scarce. The reviewed literature did not lead to conclusions regarding the potential for accelerated hydriding due to corrosion at crevices and/or second-phase particles, nor did it lead to conclusions as to the potential for a 'breakaway' in oxidation and hydrogen acquisition in long service life of Zr-Nb alloys. Specific information on service experience in U.S.S.R. power reactors could not be obtained; however, most of the information surveyed leads to the conclusion that fuel channels having Zr-2.5 wt% Nb pressure tubes should perform satisfactorily with respect to degradation from corrosion and hydriding provided they are installed correctly and are not operated under conditions that are far removed from those anticipated in design. 91 refs

  14. Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.

    1987-01-01

    This patent describes a pressurized water reactor steam generator having spaced rows of heat transfer tubes through which primary coolant from the reactor flows, the tubes being of a U-shaped design, with the U-bend portions of the U-shaped tubes stabilized by antivibration bars. The improvement described here comprises expandable antivibration bars for stabilizing the U-bend portions of the U-shaped tubes, the expandable bars having a pair of adjustable rods, formed from a pair of rod sections affixed to a connector, one rod section of each of the pair of rod sections having a plurality of protrusions. Each of the protrusions has slidable surfaces thereon. The other rod section of each of the pair of rod sections has indentations, each of the indentations having slidable surfaces thereon complementary to the sliding surfaces of the protrusions, such that the rods are expandable from a first cross-sectional width less than the spacing between two adjacent rows of the tubes, to a second cross-sectional width greater than the first cross-sectional width. The expanded rods are adapted to contact tubes of the two adjacent rows of the tubes

  15. Solvent extraction of zirconium

    International Nuclear Information System (INIS)

    Kim, S.S.; Yoon, J.H.

    1981-01-01

    The extraction of zirconium(VI) from an aqueous solution of constant ionic strength with versatic acid-10 dissolved in benzen was studied as a function of pH and the concentration of zirconium(VI) and organic acid. The effects of sulphate and chlorine ions on the extraction of the zirconium(VI) were briefly examined. It was revealed that (ZrOR 2 .2RH) is the predominant species of extracted zirconium(VI) in the versatic acid-10. The chemical equation and the apparent equilibrium constants thereof have been determined as follows. (ZrOsup(2+))aq+ 2(R 2 H 2 )sub(org) = (ZrOR 2 .2RH)sub(org)+2(H + )aq Ksub(Zr) = (ZrOR 2 .2RH)sub(org)(H + ) 2 /(ZrOsup(2+))sub(aq)(R 2 H 2 )sup(2)sub(org) = 3.3 x 10 -7 . The synergistic effects of TBP and D2EHPA were also studied. In the mixed solvent with 0.1M TBP, the synergistic effect was observed, while the mixed solvent with D2EHPA showed the antisynergistic effect. (Author)

  16. Study of diffusion processes in the oxide layer of zirconium alloys

    Directory of Open Access Journals (Sweden)

    Sialini P.

    2016-03-01

    Full Text Available In the active zone of a nuclear reactor where zirconium alloys are used as a coating material, this material is subject to various harmful impacts. During water decomposition reactions, hydrogen and oxygen are evolved that may diffuse through the oxidic layer either through zirconium dioxide (ZrO2 crystals or along ZrO2 grains. The diffusion mechanism can be studied using the Ion Beam Analysis (IBA method where nuclear reaction 18O(p,α15N is used. A tube made of zirconium alloy E110 (with 1 wt. % of Nb was used for making samples that were pre-exposed in UJP PRAHA a.s. and subsequently exposed to isotopically cleansed environment of H2 18O medium in an autoclave. The samples were analysed with gravimetric methods and IBA methods performed at the electrostatic particle accelerator Tandetron 4130 MC in the Nucler Physics Institute of the CAS, Řež. With IBA methods, the overall thicknesses of corrosion layers on the samples, element composition of the alloy and distribution of oxygen isotope 18O in the corrosion layer and its penetration in the alloy were identified. The retrieved data shows at the oxygen diffusion along ZrO2 grains because there are two peaks of 18O isotope concentrations in the corrosion layer. These peaks occur at the environment-oxide and oxide-metal interface. The element analysis identified the presence of undesirable hafnium.

  17. Purification of zirconium concentrates

    International Nuclear Information System (INIS)

    Brown, A.E.P.

    1976-01-01

    A commercial grade ZrO 2 and an ammonium uranate (yellow cake) are obtained from the caldasito ore processing. This ore is found in the Pocos de Caldas Plateau, State of Minas Gerais, Brazil. Caldasito is an uranigerous zirconium ore, a mixture of zircon and baddeleyite and contains 60% ZrO 2 and 0,3% U 3 O 8 . The chemical opening of the ore was made by alkaline fusion with NaOH at controlled temperature. The zirconium-uranium separation took place by a continuous liquid-liquid extraction in TBP-varsol-HNO 3 -H 2 O system. The raffinate containing zirconium + impurities (aluminium, iron and titanium) was purified by an ion exchange operation using a strong cationic resin [pt

  18. Determination of zirconium by fluoride ion selective electrode

    International Nuclear Information System (INIS)

    Mahanty, B.N.; Sonar, V.R.; Gaikwad, R.; Raul, S.; Das, D.K.; Prakash, A.; Afzal, Md.; Panakkal, J.P.

    2010-01-01

    Full text: Zirconium is used in a wide range of applications including nuclear clad, catalytic converters, surgical appliances, metallurgical furnaces, superconductors, ceramics, lamp filaments, anti corrosive alloys and photographical purposes. Irradiation testing of U-Zr and U-Pu-Zr fuel pins has also demonstrated their feasibility as fuel in liquid metal reactors. Different methods that are employed for the determination of zirconium are spectrophotometry, potentiometry, neutron activation analysis and mass spectrometry. Ion-selective electrode (ISE), selective to zirconium ion has been studied for the direct potentiometric measurements of zirconium ions in various samples. In the present work, an indirect method has been employed for the determination of zirconium in zirconium nitrate sample using fluoride ion selective electrode. This method is based on the addition of known excess amount of fluoride ion to react with the zirconium ion to produce zirconium tetra fluoride at about pH 2-3, followed by determination of residual fluoride ion selective electrode. The residual fluoride ion concentrations were determined from the electrode potential data using calibration plot. Subsequently, zirconium ion concentrations were determined from the concentration of consumed fluoride ions. A precision of about 2% (RSD) with the mean recovery of more than 94% has been achieved for the determination of zirconium at the concentration of 4.40 X 10 -3 moles lit -1

  19. Mitigation of harmful effects of welds in zirconium alloy components

    International Nuclear Information System (INIS)

    Coleman, C.E.; Doubt, G.L.; Fong, R.W.L.; Root, J.H.; Bowden, J.W.; Sagat, S.; Webster, R.T.

    1993-10-01

    Welding produces local residual tensile stresses and changes in texture in components made from zirconium alloys. In the heat-affected zone in tubes or plates, the basal plane normals are rotated into the plane of the component and perpendicular to the direction of the weld. Thin-walled zircaloy-2 tubes containing an axial weld do not reach their full strength, because they always fail prematurely in the weld when pressurised to failure in a fixed-end burst test. Reinforcing the weld by increasing its thickness by 25% moves the failure to the parent metal, improves the biaxial strength of the tube by 20 to 25%, and increases the total elongation by 200 to 450%. In components made from Zr-2.5Nb, the texture in the heat-affected zone promotes delayed hydride cracking (DHC) driven by tensile residual stress. Although the texture is not much affected by heat-treatments below 630 degrees celsius and large grain interaction stresses remain as a result of mixed textures, macro-residual tensile stresses can be relieved by heat-treatment to the point where the probability of cracking is very low

  20. The evaluation of validity of the RELAP5/Mod3 flow regime map for horizontal small diameter tubes at low pressure

    Energy Technology Data Exchange (ETDEWEB)

    Agafonova, N. [St. Petersburg State Technical Univ. (Russian Federation); Banati, J. [Lappeenranta Univ. of Technology (Finland)

    1997-12-31

    RELAP5/MOD3 code was developed for Western type power water reactors with vertical steam generators. Thus, this code should be validated also for WWER design with horizontal steam generators. In application for horizontal steam generators the situation with two-phase flow inside small diameter tubes is possible when the first circuit pressure drops in accident below the pressure level in the boiling water. It is known that computer codes have not always modelled correctly the two-phase flow inside horizontal tubes at low pressures (less than 4-6 MPa). It may be the result of erroneous prediction of the flow regime. Correct prediction of the flow regime is especially important for the fully or partly stratified flow in horizontal tubes. The aim of this study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal small diameter tubes. `Small diameter tube` means according RELAP5/MOD3 that the inner diameter of the tube is less (or equal) than 0.018 m. The inner tube diameter in horizontal steam generators is equal 0.013 m. (orig.). 19 refs.

  1. The evaluation of validity of the RELAP5/Mod3 flow regime map for horizontal small diameter tubes at low pressure

    Energy Technology Data Exchange (ETDEWEB)

    Agafonova, N [St. Petersburg State Technical Univ. (Russian Federation); Banati, J [Lappeenranta Univ. of Technology (Finland)

    1998-12-31

    RELAP5/MOD3 code was developed for Western type power water reactors with vertical steam generators. Thus, this code should be validated also for WWER design with horizontal steam generators. In application for horizontal steam generators the situation with two-phase flow inside small diameter tubes is possible when the first circuit pressure drops in accident below the pressure level in the boiling water. It is known that computer codes have not always modelled correctly the two-phase flow inside horizontal tubes at low pressures (less than 4-6 MPa). It may be the result of erroneous prediction of the flow regime. Correct prediction of the flow regime is especially important for the fully or partly stratified flow in horizontal tubes. The aim of this study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal small diameter tubes. `Small diameter tube` means according RELAP5/MOD3 that the inner diameter of the tube is less (or equal) than 0.018 m. The inner tube diameter in horizontal steam generators is equal 0.013 m. (orig.). 19 refs.

  2. Thermofluency in zirconium alloys

    International Nuclear Information System (INIS)

    Orozco M, E.A.

    1976-01-01

    A summary is presented about the theoretical and experimental results obtained at present in thermofluency under radiation in zirconium alloys. The phenomenon of thermofluency is presented in a general form, underlining the thermofluency at high temperature because this phenomenon is similar to the thermofluency under radiation, which ocurrs in zirconium alloys into the operating reactor. (author)

  3. Prospects for stronger calandria tubes

    International Nuclear Information System (INIS)

    Ells, C.E.; Coleman, C.E.; Hosbons, R.R.; Ibrahim, E.F.; Doubt, G.L.

    1990-12-01

    The CANDU calandria tubes, made of seam welded and annealed Zircaloy-2, have given exemplary service in-reactor. Although not designed as a system pressure containment, calandria tubes may remain intact even in the face of pressure tube rupture. One such incident at Pickering Unit 2 demonstrated the economic advantage of such an outcome, and a case can be made for increasing the probability that other calandria tubes would perform in a similar fashion. Various methods of obtaining stronger calandria tubes are available, and reviewed here. When the tubes are internally pressurized, the weld is the weak section of the tube. Increasing the oxygen concentration in the starting sheet, and thickening the weld, are promising routes to a stronger tube

  4. Reactor fuel cladding tube with excellent corrosion resistance and method of manufacturing the same

    International Nuclear Information System (INIS)

    Okuda, Takanari; Kanehara, Mitsuo; Abe, Katsuhiro; Nishimura, Takashi.

    1995-01-01

    The present invention provides a fuel cladding tube having an excellent corrosion resistance and thus a long life, and a suitable manufacturing method therefor. Namely, in the fuel cladding tube, the outer circumference of an inner layer made of a zirconium base alloy is coated with an outer layer made of a metal more corrosion resistant than the zirconium base alloy. Ti or a titanium alloy is suitable for the corrosion resistant metal. In addition, the outer layer can be coated by a method such as vapor deposition or plating, not limited to joining of the inner layer material and the outer layer material. Specifically, a composite material having an inner layer made of a zirconium alloy coated by the outer material made of a titanium alloy is applied with hot fabrication at a temperature within a range of from 500 to 850degC and at a fabrication rate of not less than 5%. The fabrication method includes any of extrusion, rolling, drawing, and casting. As the titanium-base alloy, a Ti-Al alloy or a Ti-Nb alloy containing Al of not more than 20wt%, or Nb of not more than 20wt% is preferred. (I.S.)

  5. Quantitative analysis of nickel in zirconium and zircaloy; Dosage du nickel dans le zirconium et dans le zircaloy

    Energy Technology Data Exchange (ETDEWEB)

    Rastoix, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    A rapid spectrophotometric has been developed for determination of 10 to 1000 ppm of Ni in zirconium and zircaloy using dimethylglyoxime. Iron, copper, tin and chromium, do not interfere at the concentration usually present in zirconium and its alloys. (author) [French] On determine colorimetriquenent 10 a 1000 ppm de Ni dans le zirconium et le zircaloy par photo colorimetrie a 440 m{mu} de la dimethylglyoxime nickelique. Le dosage est rapide. Le fer, le cuivre, l'etain, le chrome ne genent pas aux concentrations habituellement rencontrees dans le zirconium et ses alliages. (auteur)

  6. 40 CFR 721.9973 - Zirconium dichlorides (generic).

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 30 2010-07-01 2010-07-01 false Zirconium dichlorides (generic). 721... Substances § 721.9973 Zirconium dichlorides (generic). (a) Chemical substance and significant new uses subject to reporting. (1) The chemical substances identified generically as zirconium dichlorides (PMNs P...

  7. Molten salt scrubbing of zirconium or hafnium tetrachloride

    International Nuclear Information System (INIS)

    Lee, E.D.; McLaughlin, D.F.

    1990-01-01

    This patent describes a continuous process for removing impurities of iron or aluminum chloride or both from vaporous zirconium or hafnium chloride or both. It comprises: introducing impure zirconium or hafnium chloride vapor or both into a middle portion of an absorbing column containing a molten salt phase, the molten salt phase absorbing the impurities of iron or aluminum chloride or both to produce chloride vapor stripped of zirconium or hafnium chloride; introducing sodium or potassium chloride or both into a top portion of the column; controlling the top portion of the column to between 300--375 degrees C.; heating a bottom portion of the column to 450--550 degrees C. To vaporize zirconium chloride or hafnium chloride or hafnium and zirconium chloride from the molten salt; withdrawing molten salt substantially free of zirconium and hafnium chloride from the bottom portion of the column; and withdrawing zirconium chloride or hafnium chloride or hafnium and zirconium chloride vapor substantially free of impurities of iron and aluminum chloride from the top of the column

  8. Advances in zirconium technology for nuclear reactor application

    International Nuclear Information System (INIS)

    Ganguly, C.

    2002-01-01

    Zirconium alloys are extensively used as a material for cladding nuclear fuels and for making core structurals of water-cooled nuclear power reactors all over the world for generation of nearly 16 percent of the worlds electricity. Only four countries in the world, namely France, USA, Russia and India, have large zirconium industry and capability to manufacture reactor grade zirconium sponge, a number of zirconium alloys and a wide variety of structural components for water cooled nuclear reactor. The present paper summarises the status of zirconium technology and highlights the achievement of Nuclear Fuel Complex during the last ten years in developing a wide variety of zirconium alloys and components for water-cooled nuclear power programme

  9. Eddy current proximity measurement of perpendicular tubes from within pressure tubes in CANDU nuclear reactors

    Science.gov (United States)

    Bennett, P. F. D.; Underhill, P. R.; Morelli, J.; Krause, T. W.

    2018-04-01

    Fuel channels in CANDU® (CANada Deuterium Uranium) nuclear reactors consist of two non-concentric tubes; an inner pressure tube (PT) and a larger diameter calandria tube (CT). Up to 400 horizontally mounted fuel channels are contained within a calandria vessel, which also holds the heavy water moderator. Certain fuel channels pass perpendicularly over horizontally oriented tubes (nozzles) that are part of the reactor's liquid injection shutdown system (LISS). Due to sag, these fuel channels are at risk of coming into contact with the LISS nozzles. In the event of contact between the LISS nozzle and CT, flow-induced vibrations from within the moderator could lead to fretting and deformation of the CT. LISS nozzle proximity to CTs is currently measured optically from within the calandria vessel, but from outside the fuel channels. Measurement by an independent means would provide confidence in optical results and supplement cases where optical observations are not possible. Separation of PT and CT, known as gap, is monitored from within the PT using a transmit-receive eddy current probe. Investigation of the eddy current based gap probe as a tool to also measure proximity of LISS nozzles was carried out experimentally in this work. Eddy current response as a function of LISS-PT proximity was recorded. When PT-CT gap, PT wall thickness, PT resistivity and probe lift-off variations were not present this dependence could be used to determine the LISS-PT proximity. This method has the potential to provide LISS-CT proximity using existing gap measurement data. Obtaining LISS nozzle proximity at multiple inspection intervals could be used to provide an estimate of the time to LISS-CT contact, and thereby provide a means of optimizing maintenance schedules.

  10. Effect of a surface oxide-dispersion-strengthened layer on mechanical strength of zircaloy-4 tubes

    Directory of Open Access Journals (Sweden)

    Yang-Il Jung

    2018-03-01

    Full Text Available An oxide-dispersion-strengthened (ODS layer was formed on Zircaloy-4 tubes by a laser beam scanning process to increase mechanical strength. Laser beam was used to scan the yttrium oxide (Y2O3–coated Zircaloy-4 tube to induce the penetration of Y2O3 particles into Zircaloy-4. Laser surface treatment resulted in the formation of an ODS layer as well as microstructural phase transformation at the surface of the tube. The mechanical strength of Zircaloy-4 increased with the formation of the ODS layer. The ring-tensile strength of Zircaloy-4 increased from 790 to 870 MPa at room temperature, from 500 to 575 MPa at 380°C, and from 385 to 470 MPa at 500°C. Strengthening became more effective as the test temperature increased. It was noted that brittle fracture occurred at room temperature, which was not observed at elevated temperatures. Resistance to dynamic high-temperature bursting improved. The burst temperature increased from 760 to 830°C at a heating rate of 5°C/s and internal pressure of 8.3 MPa. The burst opening was also smaller than those in fresh Zircaloy-4 tubes. This method is expected to enhance the safety of Zr fuel cladding tubes owing to the improvement of their mechanical properties. Keywords: Laser Surface Treatment, Microstructure, Oxide Dispersion Strengthened Alloy, Tensile Strength, Zirconium Alloy

  11. Strategies to prevent ventilation-associated pneumonia: the effect of cuff pressure monitoring techniques and tracheal tube type on aspiration of subglottic secretions: an in-vitro study.

    Science.gov (United States)

    Carter, Eleanor L; Duguid, Alasdair; Ercole, Ari; Matta, Basil; Burnstein, Rowan M; Veenith, Tonny

    2014-03-01

    Ventilation-associated pneumonia (VAP) is the commonest nosocomial infection in intensive care. Implementation of a VAP prevention care bundle is a proven method to reduce its incidence. The UK care bundle recommends maintenance of the tracheal tube cuff pressure at 20 to 30  cmH₂O with 4-hourly pressure checks and use of tracheal tubes with subglottic aspiration ports in patients admitted for more than 72  h. To evaluate the effects of tracheal tube type and cuff pressure monitoring technique on leakage of subglottic secretions past the tracheal tube cuff. Bench-top study. Laboratory. A model adult trachea with simulated subglottic secretions was intubated with a tracheal tube with the cuff inflated to 25  cmH₂O. Experiments were conducted using a Portex Profile Soft Seal tracheal tube with three cuff pressure monitoring strategies and using a Portex SACETT tracheal tube with intermittent cuff pressure checks. Rate of simulated secretion leakage past the tracheal tube cuff. Mean ± SD leakage of fluid past the Profile Soft Seal tracheal tube cuff was 2.25 ± 1.49  ml  min⁻¹ with no monitoring of cuff pressure, 2.98 ± 1.63  ml  min⁻¹ with intermittent cuff pressure monitoring and 3.83 ± 2.17  ml  min⁻¹ with continuous cuff pressure monitoring (P aspiration port and aspirating the simulated secretions prior to intermittent cuff pressure checks reduced the leakage rate to 0.50 ± 0.48  ml  min⁻¹ (P aspiration port. Further evaluation of medical device performance is needed in order to design more effective VAP prevention strategies.

  12. Evaluated Plan Stress Of Weld In Pressure Tube Using X Ray Diffraction Technique

    International Nuclear Information System (INIS)

    Phan Trong Phuc; Nguyen Duc Thanh; Luu Anh Tuyen

    2011-01-01

    X ray diffraction is a fundamental technique measuring stress, this technique has determined crystal strain in materials, from that determined stress in materials. This paper presents study of evaluating plane stress of weld in pressure tube, using modern XRD apparatus: X Pert Pro. (author)

  13. A technique to simulate a tube break in a high-pressure gas/cooling water heat exchanger - HTR2008-58161

    International Nuclear Information System (INIS)

    Antwerpen, H. J. V.; Mulder, E. J.

    2008-01-01

    The gas cycles of most High Temperature Gas-Cooled Reactors (HTR's) reject heat to water at some stage. In the helium/water heat exchangers of HTR's with direct Brayton cycles, the helium is usually at a much higher pressure than the water. If the pressure boundary between the helium and the water fails inside the heat exchanger. the effect on the rest of the water system has to be established in order to do a proper system design. This can be done most efficiently by using a system simulation code, however, very few system simulation codes has the capability to do gas/liquid interface tracking as required for this problem. This study describes a calculation method with which a gas/liquid heat exchanger tube rupture can be calculated in a simulation code without interface tracking. The course of events after tube rupture is described and appropriate calculation models derived. A mathematical model for a pressure relief valve (PRV) was also created. The calculation models were implemented in the system simulation software Flownex and used to study a tube rupture on a 5000 kPa helium/water heat exchanger. The assembled calculation network solved stable and within reasonable time. The simulation provided insight into the course of events following the tube break. It was shown that the acceleration of water out of the helium cooler, by choked-flow helium, caused the main pressure pulses during the event. The maximum pressure in the water loop occurs on the opposite side of the helium cooler due to constructive interference of the initial pressure wave with itself. It was also shown that by changing only pipe lengths, the system could become prone to severe oscillations after a tube rupture event. (authors)

  14. Extraction and determination of hydrogen in uranium and zirconium

    International Nuclear Information System (INIS)

    Champeix, L.; Coblence, G.; Darras, R.

    1959-01-01

    The method of desorption under vacuum at high temperatures in the solid phase, which gives good results in the case of steels, has been applied to uranium and zirconium. In these two metals hydrogen is found mainly in the form of hydride. It is chiefly a question of determining the most suitable temperature and the heating time necessary to obtain an almost total extraction of hydrogen. Two considerations must be taken into account in the choice of temperature. It should be such that on the one hand the hydride decomposes rapidly and completely at the reduced pressure applied, and on the other hand the diffusion of hydrogen through the metal takes place fairly quickly. The apparatus and the method used are described; systematic tests have led to the adoption of temperatures of 650 deg. C for uranium and 1050 deg. C for zirconium. (author) [fr

  15. Extraction and determination of hydrogen in uranium and zirconium; Extraction et dosage de l'hydrogene dans l'uranium et le zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Champeix, L; Coblence, G; Darras, R [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    The method of desorption under vacuum at high temperatures in the solid phase, which gives good results in the case of steels, has been applied to uranium and zirconium. In these two metals hydrogen is found mainly in the form of hydride. It is chiefly a question of determining the most suitable temperature and the heating time necessary to obtain an almost total extraction of hydrogen. Two considerations must be taken into account in the choice of temperature. It should be such that on the one hand the hydride decomposes rapidly and completely at the reduced pressure applied, and on the other hand the diffusion of hydrogen through the metal takes place fairly quickly. The apparatus and the method used are described; systematic tests have led to the adoption of temperatures of 650 deg. C for uranium and 1050 deg. C for zirconium. (author) [French] La methode de desorption sous vide a chaud en phase solide, methode qui donne de bons resultats dans le cas des aciers, a ete appliquee a l'uranium et au zirconium. Dans ces deux metaux, l'hydrogene se trouve surtout a l'etat d'hydrure. Il s'agit essentiellement de determiner la temperature optimum et la duree du chauffage necessaire pour obtenir une extraction d'hydrogene pratiquement complete. Deux considerations interviennent dans le choix de la temperature. Elle doit etre telle que, d'une part la decomposition de l'hydrure se fasse rapidement et completement sous la pression reduite realisee et d'autre part que la diffusion de l'hydrogene a travers le metal soit assez rapide. L'appareil et le mode operatoire utilises sont decrits des essais systematiques ont conduit a adopter une temperature de 650 deg. C pour l'uranium et de 1050 deg. C pour le zirconium. (auteur)

  16. Bruce and Darlington power pulse and pressure tube integrity programs -status 1995

    Energy Technology Data Exchange (ETDEWEB)

    Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Wylie, J [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

    1996-12-31

    The optimum solution to pressure tube fretting at the inlet of the Bruce and Darlington channels, a concern which became very serious following inspections in early 1992, is to remove the inlet bundle and operate with a 12 fuel bundle channel. During analysis of this operating mode a `power pulse` was identified which could occur during an inlet header break where all the fuel in the channel moved rapidly to the inlet of the channel. The pulse was unacceptable and the units were derated until solutions could be implemented. A number of solutions were identified and each station has begun implementation of their specific solution. Implementation has not been without problems and this paper provides a status report on the progress to date of the long bundle implementation solution for Bruce B and Darlington and the fuelling with the flow solution being implemented at Bruce A. Both types of solution have a significant impact on the original concern, fretting of the pressure tube. (author). 1 ref., 6 figs.

  17. Condensation heat transfer and pressure drop of R-410A in a 7.0 mm O.D. microfin tube at low mass fluxes

    Science.gov (United States)

    Kim, Nae-Hyun

    2016-12-01

    R-410A condensation heat transfer and pressure drop data are provided for a 7.0 mm O.D. microfin tube at low mass fluxes (50-250 kg/m2 s). The heat transfer coefficient of the microfin tube shows a minimum behavior with the mass flux. At a low mass flux, where flow pattern is stratified, condensation induced by surface tension by microfins overwhelms condensation induced by shear, and the heat transfer coefficient decreases as mass flux increases. At a high mass flux, where flow pattern is annular, condensation induced by shear governs the heat transfer, and the heat transfer coefficient increases as mass flux increases. The pressure drop of the microfin tube is larger than that of the smooth tube at the annular flow regime. On the contrary, the pressure drop of the smooth tube is larger than that of the microfin tube at the stratified flow regime.

  18. Depressurization experiments on a plugged fibrous insulation in a horizontal pressure tube

    International Nuclear Information System (INIS)

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

    1977-08-01

    Hot gas ducts for high-temperature reactors with a helium turbine are subject to additional operational loads not caused by the gas temperature. They include vibrations, caused by high gas velocities or by the sound fields emitted from the turbine, and stresses, originating from fast, short-time pressure changes. Such pressure changes occur as a rule if the generator coupled with the turbine has to be disconnected from the grid. In order to avoid no-load operation of the turbine a bypass between HP and LP side of the turbine is opened. As a consequence of this measure a sudden pressure drop occurs in the free flow cross-section causing differential pressures within the insulation. As the size of these differential pressures depends on the insulating material, the density of plugging, the kind of internals, and on the position and size of the depressurization borings, the pressure distributions in the insulation were measured on a test tube for the HP channel. (orig./RW) [de

  19. Applications for zirconium and columbium alloys

    International Nuclear Information System (INIS)

    Condliff, A.F.

    1986-01-01

    Currently, zirconium and columbium are used in a wide range of applications, overlapping only in the field of corrosion control. As a construction material, zirconium is primarily used by the nuclear power industry. The use of zirconium in the chemical processing industry (CPI) is, however, increasing steadily. Columbian alloys are primarily applied as superconducting alloys for research particle accelerators and fusion generators as well as in magnetic resonance imaging for medical diagnosis

  20. Flow condensation pressure drop characteristics of R410A-oil mixture inside small diameter horizontal microfin tubes

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Xiangchao; Ding, Guoliang; Hu, Haitao; Zhu, Yu [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Gao, Yifeng [International Copper Association Shanghai Office, Shanghai 200020 (China); Deng, Bin [Institute of Heat Transfer Technology, Golden Dragon Precise Copper Tube Group Inc., Shanghai 200135 (China)

    2010-11-15

    Flow condensation pressure drop characteristics of R410A-oil mixture inside small diameter (5.0 mm and 4.0 mm O.D.) horizontal microfin tubes were investigated experimentally covering nominal oil concentrations from 0% to 5%. The research results indicate that, comparing with the frictional pressure drop of pure R410A, the frictional pressure drop of R410A-oil mixture may decrease by maximum of 18% when the vapor quality is lower than 0.6, and increase by maximum of 13% when the vapor quality is higher than 0.6. A new frictional pressure drop correlation for R410A-oil mixture flow condensation inside microfin tubes is developed based on the refrigerant-oil mixture properties, and can agree with 94% of the experimental data within a deviation of -30% to +30%. (author)

  1. Identification of the zirconium hydrides metallography in zircaloy-2; Contribucion al estudio por metalografia de los hidruros de circonio en Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Gonzalez, F

    1968-07-01

    Technique for the Identification of the zirconium hydrides in metallographic specimens have been developed. Microhardness, quantitative estimation and relative orientation of the present hydrides as well as grain size determination of the different Zircaloy-2 tube specimens have also been made. The specimens used were corrosion- tested in water during various periods of time at 300 degree castrating, prior to the metallographic examination. Reference specimens, as received, and heavily hydride specimens in a hydrogen atmosphere at 800 degree centigrees, have been used in the previous stages of the work. No difficulties have been met in this early stage of acquaintanceship with the zirconium hydrides. (Author) 5 refs.

  2. Development of In-Service Inspection system for heat transfer tubes in the primary pressurized water cooler in the HTTR

    Energy Technology Data Exchange (ETDEWEB)

    Shinozaki, Masayuki; Furusawa, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Wada, Shigeyuki

    1999-08-01

    The ISI (In-Service Inspection) system has been developed so as to maintain the structural integrity of heat transfer tubes in the primary pressurized water cooler in the HTTR (High Temperature Engineering Test Reactor). This system consists of eddy current probes, ultra-sonic probes, insertion and extraction units, positioning unit and so on. Verification and performance tests of the developed ISI system were carried out using mock-up heat transfer tubes in the primary pressurized water cooler. The constitution of the system, R and D results of the inspection probes, and verification and performance test results of the ISI system for heat transfer tubes are described in this paper. (author)

  3. Oxidized zirconium on ceramic; Catastrophic coupling.

    Science.gov (United States)

    Ozden, V E; Saglam, N; Dikmen, G; Tozun, I R

    2017-02-01

    Oxidized zirconium (Oxinium™; Smith & Nephew, Memphis, TN, USA) articulated with polyethylene in total hip arthroplasty (THA) appeared to have the potential to reduce wear dramatically. The thermally oxidized metal zirconium surface is transformed into ceramic-like hard surface that is resistant to abrasion. The exposure of soft zirconium metal under hard coverage surface after the damage of oxidized zirconium femoral head has been described. It occurred following joint dislocation or in situ succeeding disengagement of polyethylene liner. We reported three cases of misuse of Oxinium™ (Smith & Nephew, Memphis, TN, USA) heads. These three cases resulted in catastrophic in situ wear and inevitable failure although there was no advice, indication or recommendation for this use from the manufacturer. Copyright © 2016 Elsevier Masson SAS. All rights reserved.

  4. Evaporation heat transfer and pressure drop of R-410A in three 7.0 mm O.D. microfin tubes having different inside geometries

    International Nuclear Information System (INIS)

    Kim, Nae Hyun

    2015-01-01

    R-410A evaporation heat transfer and pressure drop data are provided for three 7.0 mm O.D. microfin tubes. The microfin tubes had different helix angle, fin height and fin apex angle. Tests were conducted for a range of quality (0.2 ∼ 0.8), mass flux (216 ∼ 390 kg/m 2 s), heat flux (9 ∼ 17 kW/m 2 ) and saturation temperature (8 ∼ 12 .deg. C). It was found that three microfin tubes yielded approximately the same heat transfer coefficients. Microfin tube with larger inter-fin spacing or smaller helix angle yielded lager pressure drop. Both heat transfer coefficient and pressure drop increased as mass flux or quality increased. However, they decreased as saturation temperature increased. The range of heat transfer enhancement factor (1.37 ∼ 1.97) was comparable with that of pressure drop penalty factor (1.22 ∼ 1.77). Data are compared with available heat transfer and pressure drop correlations

  5. Electrochemical-metallothermic reduction of zirconium in molten salt solutions

    International Nuclear Information System (INIS)

    McLaughlin, D.F.; Talko, F.

    1990-01-01

    This patent describes a method for separating hafnium from zirconium of the type wherein a feed containing zirconium and hafnium chlorides is prepared from zirconium-hafnium chloride and the feed is introduced into a distillation column, which distillation column has a reboiler connected at the bottom and a reflux condenser connected at the top and wherein a hafnium chloride enriched stream is taken from the top of the column and a zirconium enriched chloride stream is taken from the bottom of the column. It comprises: reducing the zirconium enriched chloride stream taken from the distillation column to metal by electrochemically reducing an alkaline earth metal in a molten salt bath with the molten salt in the molten salt bath consisting essentially of a mixture of at least one alkali metal chloride and at least one alkaline earth metal chloride and zirconium chloride, with the reduced alkaline earth metal reacting with the zirconium chloride to produce zirconium metal and alkaline earth metal chloride

  6. Lithium uptake and the corrosion of zirconium alloys in aqueous lithium hydroxide solutions

    International Nuclear Information System (INIS)

    Ramasubramanian, N.

    1991-01-01

    This paper reports on corrosion films on zirconium alloys that were analyzed for lithium by Atomic Absorption Spectroscopy (AAS), Secondary Ion Mass Spectrometry (SIMS), and Infrared Reflection Absorption Spectroscopy (IRAS). The oxides grown in reactor in dilute lithium hydroxide solution, specimens cut from Zircaloy, and Zr-2.5Nb alloy pressure tubes removed from CANDU (Canada Deuterium Uranium, Registered Trademark) reactors showed low concentrations of lithium (4 to 50 ppm). The lithium was not leachable in a warm dilute acid. 6 Li undergoes transmutation by the 6 Li(n,t) 4 He reaction. However, SIMS profiles for d 7 Li were identical through the bulk oxide and the isotopic ratio was close to the natural abundance value. The lithium in the oxide, existing as adsorbed lithium on the surface, has been in dynamic equilibrium with lithium in the coolant, and, in spite of many Effective Full Power Years (EFPY) of operation, lithium added to the CANDU coolant at ∼2.5 ppm is not concentrating in the oxides. On the other hand, corrosion films grown in the laboratory in concentrated lithium hydroxide solutions were very porous and contained hundreds of ppm of lithium in the oxide

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  8. 1st RCM of IAEA CRP on Prediction of Axial and Radial Creep in HWR Pressure Tubes

    International Nuclear Information System (INIS)

    Choi, Jong-Ho

    2013-01-01

    Expected outcome: Improved understanding of pressure tube creep mechanism by studying the effect of intrinsic (material response) as well as extrinsic parameters (operating conditions). • Improvement of material characterization technology: many laboratories participating in this CRP will conduct the microstructure characterization for the first time. • Recommendation for manufacturing to achieve optimal PT performance: The database will enable the identification of best pressure tube performance by comparison of data. • Improvement in aging management procedure: (channel selection for PT deformation management, etc.). • Safety enhancement for operating HWRs by reducing the uncertainty in the prediction of PT deformation

  9. Oxidation of zirconium alloys in steam: influence of tetragonal zirconia on oxide growth mechanism

    International Nuclear Information System (INIS)

    Godlewski, J.

    1990-07-01

    The oxidation of zirconium alloys in presence of steam, presents after a 'parabolic' growth law, an acceleration of the oxidation velocity. This phenomenon limits the use of zirconium alloys as nuclear fuel cladding element. In order to determine the physico-chemical process leading to this kinetic transition, two approaches have been carried out: the first one has consisted to determine the composition of the oxide layer and its evolution with the oxidation time; and the second one to determine the oxygen diffusion coefficients in the oxide layers of pre- and post-transition as well as their evolution with the oxidation time. The composition of the oxide layers has been determined by two analyses techniques: the X-ray diffraction and the laser Raman spectroscopy. This last method has allowed to confirm the presence of tetragonal zirconium oxide in the oxide layers. Analyses carried out by laser Raman spectroscopy on oxides oblique cuttings have revealed that the tetragonal zirconium oxide is transformed in monoclinic phase during the kinetic transition. A quantitative approach has allowed to corroborate the results obtained by these two techniques. In order to determine the oxygen diffusion coefficients in the oxides layers, two diffusion treatments have been carried out: 1)under low pressure with D 2 18 O 2 ) under high pressure in an autoclave with H 2 18 O. The oxygen 18 concentration profiles have been obtained by two analyses techniques: the nuclear microprobe and the secondary ions emission spectroscopy. The obtained profiles show that the mass transport is made by the volume and particularly by the grain boundaries. The corresponding diffusion coefficients have been calculated with the WHIPPLE and LE CLAIRE solution. The presence of tetragonal zirconium oxide, its relation with the kinetic transition, and the evolution of the diffusion coefficients with the oxidation time, are discussed in terms of internal stresses in the oxide layer and of the oxide layer

  10. Investigations of titamium and zirconium hydrides to determine suitability of recoverable tritium immobilization for the Pickering tritium removal system

    International Nuclear Information System (INIS)

    Noga, J.O.

    1981-11-01

    A tritium removal system will be constructed at Pickering Nuclear Generating station to reduce the adverse effects of this radioactive hydrogen isotope. This report summarizes various properties of titanium and zirconium sponge hydrides which have been selected as suitable candidates for tritium product immobilization. Equilibrium pressure-composition-temperature data indicates that both materials behave suitably to provide a safe, solid form of tritium storage. Titanium tritide is recommended as the best choice due to higher dissociation pressures which can be achieved at equivalent temperatures when compared to zirconium tritide. Higher dissociation pressures would result in faster and more efficient recovery of tritium gas from the immobilized state. It is evident from the stability of these compounds that their utilization as tritides will greatly enhance the integrity of tritium storage

  11. Laser interferometer system for the measurement of creep in pressurized tubes

    International Nuclear Information System (INIS)

    Kirchner, T.L.

    1976-07-01

    A laser interferometer measurement system was developed to measure the length, diameter, and radius of various pressurized tube specimens. The machine measures and records profilometric data of the pressurized tubes prior to insertion in the reactor and then again after a predetermined fluence has been reached to determine the amount of creep which has occurred. This data provides a statistical basis for the description of steady-state in-reactor creep and creep rupture behavior of the reference fuel cladding and structural materials for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor (CRBR). In addition, this data will be used to determine the relative in-reactor creep and creep rupture behavior of candidate alloys for advanced cladding and structural materials. The laser interferometer system, referred to as the Biaxial Creep Measurement Machine (BCMM), was built to meet or exceed design criteria such as: automatic measurement of the five biaxial creep specimens varying in size; complete automation of the machine using a mini-computer; complete specimen loading, unloading, and data processing in less than five minutes; storage of data on magnetic cassette tapes; quick-look data readout and error checking during each run to determine proper machine operation; and remote operation in a radioactive environment

  12. Effect of beta phase composition and surface machining on the oxidation behavior of Zr-2.5Nb pressure tube material

    International Nuclear Information System (INIS)

    Nouduru, S.K.; Kiran Kumar, M.; Kain, V.; Khanna, A.S.

    2015-01-01

    Zr-2.5Nb is commonly used as the pressure tube material in pressurized heavy water reactors. it is also the pressure tube material for Advanced Heavy Water Reactor (AHWR) being developed indigenously in India with light water as coolant and water chemistry similar to Boiling Water Reactors (BWR). Oxidation of the pressure tube depends on various factors like material composition, microstructure, fabrication route, and water chemistry. In the present research, the role of the composition and morphology of second phase β on the high temperature and pressure oxidation behavior of Zr-2.5Nb pressure tube material in steam was systematically studied. The as-received pressure tube material (fabricated through cold worked and stress relieved, CWSR route) was subjected to selective heat treatments to generate microstructures containing predominantly β(Zr) (∼ 20% Nb) and β(Nb) (∼ 80% Nb) phases. The presence of such phases was characterized by X-ray diffraction and transmission electron microscopy-energy dispersive spectroscopy. Subsequently both the heat treated materials were subjected to surface machining. The Zr-2.5Nb material in different microstructural conditions was subjected to accelerated oxidation exposures in steam at 400 C. degrees, and 10 MPa pressure up to 30 days. Raman spectroscopy was carried out on the oxide surfaces to observe the variation in tetragonal versus monoclinic phase fractions with oxidation duration. The microstructure consisting of predominantly β(Nb) showed a relatively improved oxidation resistance as compared to the one with predominantly β(Zr). The tetragonal phase fraction in the oxide film decreased with oxidation time in all microstructural conditions and was found to be the least in the microstructure containing β(Zr) after 10 days of exposures. The explanation for the observed higher oxidation resistance of β(Nb) microstructure lies in the context of depleted matrix Nb content in the case of β(Nb). Surface machining

  13. Problems of zirconium metal production in Czechoslovakia

    International Nuclear Information System (INIS)

    Vareka, J.; Vaclavik, E.

    1975-01-01

    The problems are summed up of the production and quality control of zirconium sponge. A survey is given of industrial applications of zirconium in form of pure metal or alloys in nuclear power production, ferrous and non-ferrous metallurgy, chemical engineering and electrical engineering. A survey is also presented of the manufacture of zirconium metal in advanced capitalist countries. (J.B.)

  14. Radiation induced defect flux behaviors at zirconium based component

    International Nuclear Information System (INIS)

    Choi, Sang Il; Kim, Ji Hyun; Kwon, Jun Hyun; Lee, Gyeong Geun

    2013-01-01

    In commercial reactor core, structure materials are located in high temperature and high pressure environment. Therefore, main concern of structure materials is corrosion and mechanical properties change than radiation effects on materials. However, radiation effects on materials become more important phenomena because research reactor condition is different from commercial reactor. The temperature is lower than 100 .deg. C and radiation dose is much higher than that of commercial reactor. Among the radiation effect on zirconium based metal, radiation induced growth (RIG), known as volume conservative distortion, is one of the most important phenomena. Recently, theoretical RIG modeling based on radiation damage theory (RDT) and balance equation are developed. However, these growth modeling have limited framework of single crystal and high temperature. To model theoretical RIG in research reactor, qualitative mechanism must be set up. Therefore, this paper intent is establishing defect flux mechanism of zirconium base metal in research reactor for RIG modeling. After than theoretical RIG work will be expanded to research reactor condition

  15. Zirconium nitride hard coatings

    International Nuclear Information System (INIS)

    Roman, Daiane; Amorim, Cintia Lugnani Gomes de; Soares, Gabriel Vieira; Figueroa, Carlos Alejandro; Baumvol, Israel Jacob Rabin; Basso, Rodrigo Leonardo de Oliveira

    2010-01-01

    Zirconium nitride (ZrN) nanometric films were deposited onto different substrates, in order to study the surface crystalline microstructure and also to investigate the electrochemical behavior to obtain a better composition that minimizes corrosion reactions. The coatings were produced by physical vapor deposition (PVD). The influence of the nitrogen partial pressure, deposition time and temperature over the surface properties was studied. Rutherford backscattering spectrometry (RBS), X-ray photoelectron spectroscopy (XPS), X-ray diffraction (XRD), scanning electron microscopy (SEM) and corrosion experiments were performed to characterize the ZrN hard coatings. The ZrN films properties and microstructure changes according to the deposition parameters. The corrosion resistance increases with temperature used in the films deposition. Corrosion tests show that ZrN coating deposited by PVD onto titanium substrate can improve the corrosion resistance. (author)

  16. Transfer of a cold atmospheric pressure plasma jet through a long flexible plastic tube

    International Nuclear Information System (INIS)

    Kostov, Konstantin G; Prysiazhnyi, Vadym; Honda, Roberto Y; Machida, Munemasa

    2015-01-01

    This work proposes an experimental configuration for the generation of a cold atmospheric pressure plasma jet at the downstream end of a long flexible plastic tube. The device consists of a cylindrical dielectric chamber where an insulated metal rod that serves as high-voltage electrode is inserted. The chamber is connected to a long (up to 4 m) commercial flexible plastic tube, equipped with a thin floating Cu wire. The wire penetrates a few mm inside the discharge chamber, passes freely (with no special support) along the plastic tube and terminates a few millimeters before the tube end. The system is flushed with Ar and the dielectric barrier discharge (DBD) is ignited inside the dielectric chamber by a low frequency ac power supply. The gas flow is guided by the plastic tube while the metal wire, when in contact with the plasma inside the DBD reactor, acquires plasma potential. There is no discharge inside the plastic tube, however an Ar plasma jet can be extracted from the downstream tube end. The jet obtained by this method is cold enough to be put in direct contact with human skin without an electric shock. Therefore, by using this approach an Ar plasma jet can be generated at the tip of a long plastic tube far from the high-voltage discharge region, which provides the safe operation conditions and device flexibility required for medical treatment. (paper)

  17. [Prehospital airway management of laryngeal tubes. Should the laryngeal tube S with gastric drain tube be preferred in emergency medicine?].

    Science.gov (United States)

    Dengler, V; Wilde, P; Byhahn, C; Mack, M G; Schalk, R

    2011-02-01

    Laryngeal tubes (LT) are increasingly being used for emergency airway management. This article reports on two patients in whom out-of-hospital intubation with a single-lumen LT was associated with massive pulmonary aspiration in one patient and gastric overinflation in the other. In both cases peak inspiratory pressures exceeded the LT leak pressure of approximately 35 mbar. This resulted in gastric inflation and decreased pulmonary compliance and increased inspiratory pressure further, thereby creating a vicious circle. It is therefore recommended that laryngeal tube suction (LTS) should be used in all cases of emergency airway management and a gastric drain tube be inserted through the dedicated second lumen. Apart from gastric overinflation, incorrect LT/LTS placement must be detected and immediately corrected, e.g. in cases of difficult or impossible gastric tube placement, permanent drainage of air from the gastric tube, decreasing minute ventilation or an ascending capnography curve.

  18. Eustachian tube patency

    Science.gov (United States)

    Eustachian tube patency refers to how much the eustachian tube is open. The eustachian tube runs between the middle ear and the throat. It controls the pressure behind the eardrum and middle ear space. This helps keep ...

  19. Measurement of subcooled boiling pressure drop and local heat transfer coefficient in horizontal tube under LPLF conditions

    International Nuclear Information System (INIS)

    Baburajan, P.K.; Bisht, G.S.; Gupta, S.K.; Prabhu, S.V.

    2013-01-01

    Highlights: ► Measured subcooled boiling pressure drop and local heat transfer coefficient in horizontal tubes. ► Infra-red thermal imaging is used for wall temperature measurement. ► Developed correlations for pressure drop and local heat transfer coefficient. -- Abstract: Horizontal flow is commonly encountered in boiler tubes, refrigerating equipments and nuclear reactor fuel channels of pressurized heavy water reactors (PHWR). Study of horizontal flow under low pressure and low flow (LPLF) conditions is important in understanding the nuclear core behavior during situations like LOCA (loss of coolant accidents). In the present work, local heat transfer coefficient and pressure drop are measured in a horizontal tube under LPLF conditions of subcooled boiling. Geometrical parameters covered in this study are diameter (5.5 mm, 7.5 mm and 9.5 mm) and length (550 mm, 750 mm and 1000 mm). The operating parameters varied are mass flux (450–935 kg/m 2 s) and inlet subcooling (29 °C, 50 °C and 70 °C). Infra-red thermography is used for the measurement of local wall temperature to estimate the heat transfer coefficient in single phase and two phase flows with water as the working medium at atmospheric pressure. Correlation for single phase diabatic pressure drop ratio (diabatic to adiabatic) as a function of viscosity ratio (wall temperature to fluid temperature) is presented. Correlation for pressure drop under subcooled boiling conditions as a function of Boiling number (Bo) and Jakob number (Ja) is obtained. Correlation for single phase heat transfer coefficient in the thermal developing region is presented as a function of Reynolds number (Re), Prandtl number (Pr) and z/d (ratio of axial length of the test section to diameter). Correlation for two phase heat transfer coefficient under subcooled boiling condition is developed as a function of boiling number (Bo), Jakob number (Ja) and Prandtl number (Pr)

  20. Evaluation of maintenance strategies for steam generator tubes in pressurized waster reactors. 2. Cost and profitability analyses

    International Nuclear Information System (INIS)

    Isobe, Y.; Sagisaka, M.; Yoshimura, S.; Yagawa, G.

    2000-01-01

    As an application of probabilistic fracture mechanics (PFM), risk-benefit analysis was carried out to evaluate maintenance activities of steam generator (SG) tubes used in pressurized water reactors (PWRs). The analysis was conducted for SG tubes made of Inconel 600, and Inconel 690 as well assuming its crack initiation and crack propagation law based on Inconel 600 data. The following results were drawn from the analysis. Improvement of inspection accuracy reduces the maintenance costs significantly and is preferable from the viewpoint of profitability due to reduction of SG tube leakage and rupture. There is a certain region of SCC properties of SG tubes where sampling inspection is effective. (author)

  1. CANDU pressure tube leak detection by annulus gas dew point measurement. A critical review

    International Nuclear Information System (INIS)

    Greening, F.R.

    2017-01-01

    In the event of a pressure tube leak from a small through-wall crack during CANDU reactor operations, there is a regulatory requirement - referred to as Leak Before Break (LBB) - for the licensee to demonstrate that there will be sufficient time for the leak to be detected and the reactor shut down before the crack grows to the critical size for fast-uncontrolled rupture. In all currently operating CANDU reactors, worldwide, this LBB requirement is met via continuous dew point measurements of the CO_2 gas circulating in the reactor's Annulus Gas System (AGS). In this paper the historical development and current status of this leak detection capability is reviewed and the use of moisture injection tests as a verification procedure is critiqued. It is concluded that these tests do not represent AGS conditions that are to be expected in the event of a real pressure tube leak.

  2. CANDU pressure tube leak detection by annulus gas dew point measurement. A critical review

    Energy Technology Data Exchange (ETDEWEB)

    Greening, F.R. [CTS-NA, Tiverton, ON (Canada)

    2017-03-15

    In the event of a pressure tube leak from a small through-wall crack during CANDU reactor operations, there is a regulatory requirement - referred to as Leak Before Break (LBB) - for the licensee to demonstrate that there will be sufficient time for the leak to be detected and the reactor shut down before the crack grows to the critical size for fast-uncontrolled rupture. In all currently operating CANDU reactors, worldwide, this LBB requirement is met via continuous dew point measurements of the CO{sub 2} gas circulating in the reactor's Annulus Gas System (AGS). In this paper the historical development and current status of this leak detection capability is reviewed and the use of moisture injection tests as a verification procedure is critiqued. It is concluded that these tests do not represent AGS conditions that are to be expected in the event of a real pressure tube leak.

  3. Phase Transformations in a Uranium-Zirconium Alloy containing 2 weight per cent Zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Lagerberg, G

    1961-04-15

    The phase transformations in a uranium-zirconium alloy containing 2 weight percent zirconium have been examined metallographically after heat treatments involving isothermal transformation of y and cooling from the -y-range at different rates. Transformations on heating and cooling have also been studied in uranium-zirconium alloys with 0.5, 2 and 5 weight per cent zirconium by means of differential thermal analysis. The results are compatible with the phase diagram given by Howlett and Knapton. On quenching from the {gamma}-range the {gamma} phase transforms martensitically to supersaturated a the M{sub S} temperature being about 490 C. During isothermal transformation of {gamma} in the temperature range 735 to 700 C {beta}-phase is precipitated as Widmanstaetten plates and the equilibrium structure consists of {beta} and {gamma}{sub 1}. Below 700 C {gamma} transforms completely to Widmanstaetten plates which consist of {beta} above 660 C and of a at lower temperatures. Secondary phases, {gamma}{sub 2} above 610 C and {delta} below this temperature, are precipitated from the initially supersaturated Widmanstaetten plates during the isothermal treatments. At and slightly below 700 C the cooperative growth of |3 and {gamma}{sub 2} is observed. The results of isothermal transformation are summarized in a TTTdiagram.

  4. The influence of the preliminary garter spring spacer simulator clamping force in the pressure tube spacer -calandria tube hook-up simulator aging behaviour

    International Nuclear Information System (INIS)

    Gyongyosi, T.; Deloreanu, G.; Puiu, D.; Corbescu, B.; Anghel, N.; Dinu, E.

    2016-01-01

    The garter spring spacer is a specially constructed torsion spring used to fit-out the CANDU 6 fuel channel. The pressure tube ageing decreases the gap to the calandria tube. Continuous gap decrease directly affects the garter spring spacers behavior during fuel channel assembly operation. The preliminary clamping force value of the garter spring spacer assembly is important for its ageing behavior. This paper briefly describes the experimental technological facilities used for conducted the experiments and highlights some of the important moments during an experiment carried out in laboratory conditions, without using pressurized boiled water and irradiation working conditions. The results analysis and some conclusions are outlined at the end, pointing out that a garter spring spacer preliminary clamping force increase reduces the vibration response signal amplitude, and does not lead to its relaxation. The paper is dedicated to specialists working in research and technological engineering. (authors)

  5. Comparison of endotracheal tube cuff pressure values before and after training seminar.

    Science.gov (United States)

    Özcan, Ayça Tuba Dumanlı; Döğer, Cihan; But, Abdülkadir; Kutlu, Işık; Aksoy, Şemsi Mustafa

    2018-06-01

    It is recommended that endotracheal cuff (ETTc) pressure be between 20 and 30 cm H 2 O. In this present study, we intend to observe average cuff pressure values in our clinic and the change in these values after the training seminar. The cuff pressure values of 200 patients intubated following general anesthesia induction in the operating theatre were measured following intubation. One hundred patients whose values were measured before the training seminar held for all physician assistants, and 100 patients whose values were measured after the training seminar were regarded as Group 1 and Group 2, respectively. Cuff pressures of both groups were recorded, and the difference between them was shown. Moreover, cuff pressure values were explored according to the working period of the physician assistants. There was no significant difference between the groups in terms of age, gender and tube diameters. Statistically significant difference was found between cuff pressure values before and after the training (p values decreased, however no statistically significant different was found (p values and potential complications.

  6. Proposal for a Coordinated Research Project: Prediction of Axial and Radial Creep in Pressure Tubes

    International Nuclear Information System (INIS)

    Bozzano, Patricia B.

    2013-01-01

    Participation of Argentina: • hydrogen charge of 90 samples for hydrogen determination: IGF, HVEMS, DSC, Resistivity, DTA; • 17 samples for non destructive techniques; • 6 blisters in CANDU pressure tube sections for NDT evaluation

  7. Method to precipitate metals on zirconium articles without using current

    International Nuclear Information System (INIS)

    Donaghy, R.E.

    1978-01-01

    Tubes or other bodies made of zirconium or Zr alloys used for taking up nuclear fuels are coated without current with a metal film to improve the mechanical and corrosion properties. The article is activated in a solution of ammonium bifluoride and sulphuric acid where an electrical conducting fixed and loose layer are fermed. This loose film is removed chemically by reacting with fluoroboric acid or hydrofluorosilica acid by ultrasonic treatment or by stripping from organic material (cotton, polyester, nylon). The current-free plating with Cu and Ni is described. The article is washed with deionized water between the process steps and finally degassed at a temperature of 150-200 0 C. (IHOE) [de

  8. The fluorimetric titration of zirconium in the ppm-range

    International Nuclear Information System (INIS)

    Linden, W.E. von der; Boef, G. den; Ozinga, W.

    1976-01-01

    A fluorimetric titration of zirconium(IV) with EDTA is proposed. The fluorescence intensity of the zirconium-morin complex is used to indicate the end-point. More than twenty other cations were investigated and it was found that they did not interfere, neither did common anions. Mercury(II) can only be tolerated in amount not exceeding that of zirconium. Bismuth(III) interferes and hafnium(IV0 is titrated together with zirconium. The relative standard deviation of the titration of 10ml of a solution containing 1 ppm of zirconium does not exceed 1.5%

  9. Zirconium behaviour during electrorefining of actinide-zirconium alloy in molten LiCl-KCl on aluminium cathodes

    Energy Technology Data Exchange (ETDEWEB)

    Meier, R. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Heidelberg University, Institute of Physical Chemistry, Im Neuenheimer Feld 253, Heidelberg 69120 (Germany); Souček, P., E-mail: Pavel.Soucek@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Malmbeck, R.; Krachler, M.; Rodrigues, A.; Claux, B.; Glatz, J.-P. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Fanghänel, Th. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, Karlsruhe 76125 (Germany); Heidelberg University, Institute of Physical Chemistry, Im Neuenheimer Feld 253, Heidelberg 69120 (Germany)

    2016-04-15

    A pyrochemical electrorefining process for the recovery of actinides from metallic nuclear fuel based on actinide-zirconium alloys (An–Zr) in a molten salt is being investigated. In this process actinides are group-selectively recovered on solid aluminium cathodes as An–Al alloys using a LiCl–KCl eutectic melt at a temperature of 450 °C. In the present study the electrochemical behaviour of zirconium during electrorefining was investigated. The maximum amount of actinides that can be oxidised without anodic co-dissolution of zirconium was determined at a selected constant cathodic current density. The experiment consisted of three steps to assess the different stages of the electrorefining process, each of which employing a fresh aluminium cathode. The results indicate that almost a complete dissolution of the actinides without co-dissolution of zirconium is possible under the applied experimental conditions. - Highlights: • Recovery of actinides was shown by electrorefining of U/Pu–Zr alloys in LiCl–KCl. • Constant current density of 20 mA/cm{sup 2} is applied. • Most of the actinides were dissolved avoiding zirconium co-dissolution. • Deterioration of the deposit quality by a small amount of co-deposited Zr is not observed.

  10. Visual inspection technology of the narrow and small confined area for monitoring feederpipe support of pressure tube in calandria reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Wan; Lee, Nam Ho; Choi, Young Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    There are 760 feederpipes, which they are connected to inlet/outlet of the 380 pressure tube channels on the front of the calandria, in CANDU-type Reactor of Wolsung Nuclear Power Plant. As an ISI(In-Service Inspection) and PSI (Post-Service Inspection) requirements, maintenance activities of measuring the thickness of curvilinear part of feederpipe and inspecting the feederpipe support area within calandria are needed to ensure continued reliable operation of nuclear power plant. And ultrasonic probe is used to measure the thickness of curvilinear part of feederpipe, however workers are exposed to radioactivity irradiation during the measurement period. But, it is exposed to radioactivity irradiation during the measurement period. But, it is impossible to inspect feederpipe support area thoroughly because of narrow and confined accessibility, that is , an inspection space between the pressure tube channels is less than 100 mm and pipes in feederpipe support area are congested. And also, workers involved in inspecting feederpipe support area are under the jeopardy of high-level radiation exposure. Concerns about sliding home, which make the move of feederpipe connected to pressure tube channel smooth as pressure tube expands and contracts in its axial direction, stuck to feedeerpipe support and some of the structural components have made necessary the development of video inspection probe system with narrow and confined accessibility to observe and inspect feederpipe support area more close. Using video inspection probe system, it is possible to inspect and repair abnormality of feederpipe support connected to pressure tube channels of the calandria more accurate and quantative than naked eye. Therefore, that will do much for ensuring safety of CANDU-type nuclear power plant. 45 figs.,31 tabs. (Author)

  11. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L.

    1997-01-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis

  12. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  13. SEPARATING HAFNIUM FROM ZIRCONIUM

    Science.gov (United States)

    Lister, B.A.J.; Duncan, J.F.

    1956-08-21

    A dilute aqueous solution of zirconyl chloride which is 1N to 2N in HCl is passed through a column of a cation exchange resin in acid form thereby absorbing both zirconium and associated hafnium impurity in the mesin. The cation exchange material with the absorbate is then eluted with aqueous sulfuric acid of a O.8N to 1.2N strength. The first portion of the eluate contains the zirconium substantially free of hafnium.

  14. Flow and pressure drop fluctuations in a vertical tube subject to low frequency oscillations

    International Nuclear Information System (INIS)

    Pendyala, Rajashekhar; Jayanti, Sreenivas; Balakrishnan, A.R.

    2008-01-01

    Heat transfer and other equipment mounted on off-shore platforms may be subjected to low frequency oscillations. The effect of these oscillations, typically in the frequency range of 0.1-1 Hz, on the flow rate and pressure drop in a vertical tube has been studied experimentally in the present work. A 1.75 m-long vertical tube of inner diameter 0.016 m was mounted on a plate and the whole plate was subjected to oscillations in the vertical plane using a mechanical simulator capable of providing low frequency oscillations in the range of 8-30 cycles/min at an amplitude of 0.125 m. The effect of the oscillations on the flow rate and the pressure drop has been measured systematically in the Reynolds number range 500-6500. The induced flow rate fluctuations were found to be dependent on the Reynolds number with stronger fluctuations at lower Reynolds numbers. The effective friction factor, based on the mean pressure drop and the mean flow rate, was also found to be higher than expected. Correlations have been developed to quantify this Reynolds number dependence

  15. Flow and pressure drop fluctuations in a vertical tube subject to low frequency oscillations

    Energy Technology Data Exchange (ETDEWEB)

    Pendyala, Rajashekhar; Jayanti, Sreenivas [Department of Chemical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Balakrishnan, A.R. [Department of Chemical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India)], E-mail: arbala@iitm.ac.in

    2008-01-15

    Heat transfer and other equipment mounted on off-shore platforms may be subjected to low frequency oscillations. The effect of these oscillations, typically in the frequency range of 0.1-1 Hz, on the flow rate and pressure drop in a vertical tube has been studied experimentally in the present work. A 1.75 m-long vertical tube of inner diameter 0.016 m was mounted on a plate and the whole plate was subjected to oscillations in the vertical plane using a mechanical simulator capable of providing low frequency oscillations in the range of 8-30 cycles/min at an amplitude of 0.125 m. The effect of the oscillations on the flow rate and the pressure drop has been measured systematically in the Reynolds number range 500-6500. The induced flow rate fluctuations were found to be dependent on the Reynolds number with stronger fluctuations at lower Reynolds numbers. The effective friction factor, based on the mean pressure drop and the mean flow rate, was also found to be higher than expected. Correlations have been developed to quantify this Reynolds number dependence.

  16. Spectrophotometric titration of zirconium in siliceous materials

    International Nuclear Information System (INIS)

    Sugawara, K.F.; Su, Y.-S.; Strzegowski, W.R.

    1978-01-01

    An accurate and selective complexometric titration procedure based upon a spectrophotometrically detected end-point has been developed for the determination of zirconium in glasses, glass-ceramics and refractories. A p-bromomandelic acid separation step for zirconium imparts excellent selectivity to the procedure. The method is particularly important for the 1 to 5% concentration range where a simple, accurate and selective method for the determination of zirconium has been lacking. (author)

  17. Wear and chemistry of zirconium-silicate, aluminium-silicate and zirconium-aluminium-silicate glasses in alkaline medium

    International Nuclear Information System (INIS)

    Rouse, C.G.; Lemos Guenaga, C.M. de

    1984-01-01

    A study of the chemical durability, in alkaline solutions, of zirconium silicate, aluminium silicate, zirconium/aluminium silicate glasses as a function of glass composition is carried out. The glasses were tested using standard DIN-52322 method, where the glass samples are prepared in small polished pieces and attacked for 3 hours in a 800 ml solution of 1N (NaOH + NA 2 CO 3 ) at 97 0 C. The results show that the presence of ZrO 2 in the glass composition increases its chemical durability to alkaline attack. Glasses of the aluminium/zirconium silicate series were melted with and without TiO 2 . It was shown experimentally that for this series of glasses, the presence of both TiO 2 and ZrO 2 gave better chemical durability results. However, the best overall results were obtained from the simpler zirconium silicate glasses, where it was possible to make glasses with higher values of ZrO 2 . (Author) [pt

  18. Study for the chlorination of zirconium oxide

    International Nuclear Information System (INIS)

    Seo, E.S.M.; Takiishi, H.; Paschoal, J.O.A.; Andreoli, M.

    1990-12-01

    In the development of new ceramic and metallic materials the chlorination process constitutes step in the formation of several intermediate compounds, such as metallic chlorides, used for the production of high, purity raw materials. Chlorination studies with the aim of fabrication special zirconium-base alloys have been carried out at IPEN. Within this program the chlorination technique has been used for zirconium tetrachloride production from zirconium oxide. In this paper some relevant parameters such as: time and temperature of reaction, flow rate of chloride gas and percentage of the reducing agent which influence the efficiency of chlorination of zirconium oxide are evaluated. Thermodynamical aspects about the reactions involved in the process are also presented. (author)

  19. Ductile zirconium powder by hydride-dehydride process

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, T S [BHABHA ATOMIC RESEARCH CENTRE, BOMBAY (INDIA); CHAUDHARY, S [NUCLEAR FUEL COMPLEX, HYDERABAD (INDIA)

    1976-09-01

    The preparation of ductile zirconium powder by the hydride-dehydride process has been described. In this process massive zirconium obtained from Kroll reduction of ZrCl/sub 4/ is first rendered brittle by hydrogenation and the hydride crushed and ground in a ball mill to the required particle size. Hydrogen is then hot vacuum extracted to yield the metal powder. The process has been successfully employed for the production of zirconium powders with low oxygen content and having hardness values in the range of 115-130 BHN, starting from a zirconium sponge of 100-120 BHN hardness. Influence of surface characteristics of the starting metal on its hydriding behaviour has been studied and the optimum hydriding-dehydriding conditions established.

  20. Joint titrimetric determination of zirconium and hafnium

    International Nuclear Information System (INIS)

    Vazquez, Cristina; Botbol, Moises; Bianco de Salas, G.N.; Cornell de Casas, M.I.

    1980-01-01

    A method for the joint titrimetric determination of zirconium and hafnium, which are elements of similar chemical behaviour, is described. The disodic salt of the ethylendiaminetetracetic acid (EDTA) is used for titration, while xilenol orange serves as final point indicator. Prior to titration it is important to evaporate with sulfuric acid, the solution resulting from the zirconium depolymerization process, to adjust the acidity and to eliminate any interferences. The method, that allows the quick and precise determination of zirconium and hafnium in quantities comprised between 0.01 and mg, was applied to the analysis of raw materials and of intermediate and final products in the fabrication of zirconium sponge and zircaloy. (M.E.L.) [es