International Nuclear Information System (INIS)

Zircaloy-2 and Zircaloy-4 tubing materials were irradiated with 1 MeV proton at 350 degrees C to doses of 0.01, 0.1, and 1 dpa respectively. Both microstructure examination and nodular corrosion test (500 degrees C, 1500 psi steam) were performed in order to understand the relationship between the microstructural evolution and the corrosion resistance of these alloys under irradiation. Neutron-irradiated Zircaloy-2 specimens which were obtained from a failed BWR fuel rod cladding were also studied. Specimens of three different neutron fluences were investigated; namely, 2.6x10"2"4, 3.2x10"2"5, 3.8x10"2"5, (E_n#>=#1MeV). The results indicated that the higher the irradiation dose the better the nodular corrosion resistance of both Zircaloy-2 and Zircaloy-4. It is concluded that irradiation-induced precipitate dissolution and ...

Energy Technology Data Exchange (ETDEWEB)

The use of nickel alloys in Flue Gas Desulfurization (FGD) applications is described. Recently, nickel alloys 625 and C-276 have been clad to carbon steel plate. Production and properties of the clad plate are reviewed and a brief discussion is presented on production, fabrication and welding. Laboratory and field corrosion tests in SO/sub 2/ scrubber environments suggest that alloy 625 and C-276 clad steel plates are as corrosion resistant as the solid alloy plate.

Energy Technology Data Exchange (ETDEWEB)

Experiments have been carried out in water at 355 C to study transport of oxygen and hydrogen (as deuterium) in growing corrosion films. Composition of the films was also examined in 2.9 Mev and 3.9 Mev /alpha/-particle backscattering experiments. Corrosion occurs predominantly by oxygen diffusion through the film via grain boundary or similar short circuit diffusion paths, to form fresh oxide at the oxide metal interface. Increasing grain size within thick pre-breakaway films contributes to a decrease in diffusivity. The rate transition results from the generation of new diffusion pathways in previously protective oxide. Unexpectedly high concentrations of deuterium were observed. 26 refs.

International Nuclear Information System (INIS)

An x-ray fluorescence spectrometer was designed and fabricated which nondestructively determines the thickness of aluminum cladding at small suspected thin spots in the inner or outer surface of actinide reactor tubes. The analysis method is based on the difference in absorption of actinide L/sub #alpha#/ and L/sub #beta#/ fluorescent x-rays in passing through the cladding. Calibration plots of the logarithm of the L/sub #beta#//L/sub #alpha#/ x-ray intensity ratio versus cladding thickness are linear to at least 40 mils for U-Al, U_3O_8-Al, and PuO_2-Al substrates. Accuracy and precision of the experimentally determined cladding thickness and evaluated for both uranium and plutonium substrates. Experimental thickness data are reported for 618 quality assurance analyses on six Mark 41 PuO_2-Al target tubes. An x-ray fluorescence cladding thickness monitor operated with a ...

International Nuclear Information System (INIS)

Approximately 8% of the spent nuclear fuel owned by the US Department of Energy is clad with aluminum alloys. The spent fuel must be either reprocessed or temporarily stored in wet or dry storage systems until a decision is made on final disposition in a repository. There are corrosion issues associated with the aluminum cladding regardless of the disposition pathway selected. This paper discusses those issues and provides data and analysis to demonstrate that control of corrosion induced degradation in aluminum clad spent fuels can be achieved through relatively simple engineering practices.

International Nuclear Information System (INIS)

Axial offset anomaly (AOA) is caused by the deposition of crud on the fuel cladding of a PWR. When significant levels of crud build up on the cladding, boron can accumulate in the pores of the crud as a concentrated solution or solid phase, and cause the flux depression. Numerous studies have been conducted on the primary water chemistry to reduce the amount of crud in the primary circuit to avoid radioactivity buildup and unexpected power transition in the plant. However, experiments on the crud are restricted in the laboratory because the crud is a highly radioactive material. The objective of this study is to develop a test method for simulating the deposition of crud in a nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

An Fe-19Cr-8Al-0.5Hf alloy has been roll bonded to Alloy 800 to form a protective clad material suitable for use in coal conversion atmospheres. The commercial producibility of both the Fe-19Cr-8Al-0.5Hf alloy and the clad composite has been demonstrated. The clad composite has been shown to exhibit resistance to sulfidation in high sulfur, coal gasification atmospheres low Btu for 2000 hrs at 1800/sup 0/F (1255K). The FeCrAlHf alloy must be preoxidized in the absence of sulfur to form fully protective alumina scales on all surfaces. Useful life of the composite at temperatures of 1400 to 2000/sup 0/F (1033 to 1366K) is governed by reactions at the FeCrAlHf clad alloy 800 interface. The rate of interdiffusion decreases with increasing Ni content and a high Ni alloy such as RA333 appears to be a better substrate than Alloy 800 for a FeCrAlHf composite. The mechanical properties of the ...

Energy Technology Data Exchange (ETDEWEB)

This report documents the technical development for a patent disclosure of a double-clad process for the continuous casting of thin-strip carbon steel. The fundamental idea of the disclosure is to form a product strip by depositing molten steel between two, cooled, clad strips of the same material. The claimed benefits include: (a) the conservation of energy in steel making through the elimination of soaking pits and reheat cycles, and (b) an improved surface on both sides of the as-cast product such that it will be suitable for direct feed to a cold-reduction mill. However, the process as conceived is not necessarily limited to the casting of carbon steel, but may be also applied to other metals and alloys. The work is described under three headings as follows. Preliminary Considerations and Scoping Analysis presents the basic idea of the double-clad, thin-strip casting process; the energy conservation potential; scoping ...

Energy Technology Data Exchange (ETDEWEB)

This report presents the results of cladding technology development under the Coal-Fired Combined Cycle DOE Project for the period July 1976 through June 1978. Sub-task 3.1, Clad Alloy Development, involved the selection, fabrication and burner rig evaluation of advanced clad alloy compositions diffusion-bonded to IN-738. Testing was conducted for up to 5800 hours at 1600/sup 0/F in a simulated PFB environment. Metallographic evaluation showed Co-base cladding alloys PFB-5PM and PFB-6PM (aluminided) and Fe-base alloy GE-2541 to be the most corrosion resistant. The formability and fabricability of sheet material was significantly improved for these alloys by utilizing powder metallurgy techniques. Sub-task 3.2, Clad Process Development, encompassed processing activities to evaluate sheet alloy formability, bucket surface preparation techniques, platform clad ...

Energy Technology Data Exchange (ETDEWEB)

Total hemispherical emittance was measured for the SP-100 reactor fuel cladding alloy (Nb-l% Zr). Based on a standard test method (ASTM C 835-82), experiments were conducted on a reference sample of oxidized stainless steel and then on a sample of actual cladding. The sample is heated in a vacuum by passing DC current through it until reaching equilibrium. Measurements are made of the electrical power dissipated in the sample and of the surface temperature. Using the Stefan-Boltzmann Law and some key assumptions concerning conductive and radiative heat transfer, the measured quantities are used to calculate emittance. Calculated values for unoxidized cladding range from 0.159 +/- 5.35% at 913 K to 0.200 +/- 4.51% at 1091 K. Highest value measured after onset of visible oxidation was 0.339 +/- 3.92% at 1269 K.... SP-100, Reactor, Emittance, Niobium, Fuel cladding, Emissivity.

Energy Technology Data Exchange (ETDEWEB)

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and ...

Science.gov (United States)

Following studies on fiber and sheet texture of hexagonal crystal system in 1988, work has been extended to tube texture. Using the zircaloy-4 fuel cladding of Wolsung-type reactor as specimen, six pole figures for different crystallographic planes were m...

Energy Technology Data Exchange (ETDEWEB)

Information is presented concerning GCFR design and limiting faulted events; selection of PC-5 temperature limit; and verification test programs.

International Nuclear Information System (INIS)

The applicability of the thermal response of an electrically heated simulated rod mostly used in loss-of-coolant-accident (LOCA) experiments to that of a nuclear fuel rod is a concern for the safety evaluation of a reactor. The present analysis describes the characteristics of the thermal response for both electrically heated and nuclear fuel rods during typical reflood conditions for a PWR-LOCA. A model describing the radial temperature field in the rod is developed based on the scheme in HETRAP code by Malang and incorporated into a reflood analysis code, REFLA for that purpose. The calculations applied to the existing reflood tests gave good agreement with experiments, showing the validity of the present model. The analysis has shown that the nuclear fuel rod tends to give a lower clad temperature and a sooner quench time than the electrically heated rod in a typical reflood condition, due to the smaller gap heat transfer and smaller heat capacity of the ...

Energy Technology Data Exchange (ETDEWEB)

This study does not recommend the use of a pressure vessel made of integral stainless steel, due to the failure mechanisms under stress corrosion assisted by chlorides or polythionic acid. Are presented case studies of literature and analysis of reports of proceedings of RLAM reactors, showing that the materials produced by bimetallic clad overlay or are more appropriate, in terms of integrity, for use in equipment that the internal environment requires austenitic stainless steel specification.

International Nuclear Information System (INIS)

Non-destructive Evaluation (NDE) plays a major role in ensuring the safe and reliable operation of PHWRs which are the mainstay of India's nuclear power programme. An important in-service inspection (ISI) requirement in these reactors is carried out through Eddy Current Testing (ECT) of the pressure tube (PT)/calandria tube (CT) assemblies. The material of construction of these assemblies is zircaloy-2. The two main objectives of this ISI are the detection of garter spring between CT and PT and the profiling of gap between CT and PT. The paper discusses the work carried out at the authors' laboratory on the development of ECT probes for ISI of PT/CT assemblies. Emphasis has been given on the work done on the design and optimisation of the probes using computer modeling. A 2-D finite element code has been developed for this purpose. The code is developed around a diffusion equation which can be derived from Maxwell's equations governing the ...

International Nuclear Information System (INIS)

Secondary ion mass spectrometry (SIMS) has been used to investigate changes that occur in an advanced Ir-0.3W alloy during high temperature aging. This alloy is used to clad "2"3"8PuO_2 heat sources used in thermoelectric generators for deep space reconnaissance satellites. Long-term direct contact with PuO_2 at 1400"0C leads to physical and chemical changes within the cladding alloy that affect its metallurgical properties. SIMS was used to show that Cr, Fe, Ni, and in some cases O, diffuse from the PuO_2 into the alloy. Thorium and aluminum diffuse out of the alloy in these same regions. This SIMS study suggests that inward O diffusion and subsequent formation of ThO_2 on grain boundaries may stabilize the alloy against enhanced grain growth.

International Nuclear Information System (INIS)

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 ...

Energy Technology Data Exchange (ETDEWEB)

The characterization data obtained to date are described for Approved Testing Material (ATM)-103, which is spent fuel from Assembly D101 of pressurized-water reactor Calvert Cliffs, No. 1. This report is one in a series being written by the Materials Characterization Center (MCC) at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US nuclear waste repository program. ATM-103 consists of 176 full-length irradiated fuel rods with rod-average burnups of about 2600 GJ/kgM (30 MWd/kgM) and less than 1% fission gas release. Characterization data include 1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; 2) isotopic gamma scans; 3) fission gas analyses; 4) ceramography of the fuel and metallography of the cladding; 5) special fuels studies involving analytical transmission electron microscopy (AEM); 6) calculated ...

International Nuclear Information System (INIS)

The super simulator 'SAMPSON' has been developed to show that there exist certain safety margins for light water reactors under hypothetical severe accidents and to investigate realistic measures of accident management by simulating accidents with a parallel computer. Heat-up of fuel rods and release of fission products from fuels are important factors to evaluate source terms. Models for fuel rod heat-up, hydrogen production due to cladding oxidation and cladding deformation and failure in the core region have been developed in the fuel rod heat-up analysis module. Fuel temperatures were calculated by solving the heat conduction equation. The calculated results for fuel temperature and hydrogen production were compared with CORA-13 experiment results. The comparisons showed prediction capability for the heat-up of fuel rods. The fission product release analysis module incorporates with models for fission product transport within fuel pellets, ...

Energy Technology Data Exchange (ETDEWEB)

A total of 86 experimental fuel pins with various fuel, liner, and cladding candidate materials have been irradiated in the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF) reactor as part of the SP-100 fuel pin irradiation testing program. Postirradiation examination results from these fuel pin are key in establishing performance correlations and demonstrating the lifetime and safety of the reactor fuel system. This paper provides a brief description of the in-reactor fuel pin tests and presents the most recent irradiation data on the performance of wrought rhenium (Re) liner material and high density UN fuel at goal burnup of 6 atom percent (at. %). It also provides an overview of the significant variety of other fuel/liner/cladding combinations which were irradiated as part of this program and which may be of interest to more advanced efforts.

International Nuclear Information System (INIS)

Ppercase(femaxi-iv) is a computer code for the analysis of thermal and mechanical behavior of light water reactor fuel rods during steady-state and transient conditions. The main purpose of ppercase(femaxi-iv) is to calculate the stress and strain distributions in the fuel and cladding due to the pellet-cladding mechanical interaction, and the fission gas release rate during operations, especially power transients. The capability of ppercase(femaxi-iv) was extensively tested with a number of experimental results obtained in some international fuel irradiation programs. This paper provides a general description of the various models involved in ppercase(femaxi-iv) and some results of applications on the experimental data. ((orig.)).

International Nuclear Information System (INIS)

The SCDAP computer code, which is being developed under the sponsorship of the United States Nuclear Regulatory Commission, models the progression of light water reactor core damage including core heatup, core disruption and debris formation, debris heatup, and debris melting. SCDAP is being used to help identify and understand the phenomena that control core behavior during a severe accident, to help quantify uncertainties in risk assessment analysis, and to support planning and interpretation of severe fuel damage experiments and data. Comparisons between SCDAP calculations and the experimental data showed good agreement. Calculated and measured bundle temperatures for SFD-ST were within 200 K for the entire bundle and within 20 K for maximum cladding temperatures. For ESSI-2, calculated and measured maximum cladding temperatures were within 50 K, and the extensive liquefaction and relocation that was calculated was in agreement with ...

Science.gov (United States)

Recent radiochemical investigations of the PWR reactor coolant have corfirmed earlier observations that the level of activities of 33 m Cs/sup 138/, 2.8 hr Kr , and 8.1 day 1/sup 131/ are more than ten times higher than those predicted for the estimated U contamination of the Zircaloy cladding. The present fission product activity levels have not, as yet, presented any problems in the PWR. (W.L.H.)

International Nuclear Information System (INIS)

HEDL is conducting an experimental transient testing program to evaluate the performance of prototypic Fast Flux Test Facility (FFTF) fuel pins up to the cladding integrity limit. The relationship is described of the HEDL/TREAT transient overpower test program to the confirmation of FFTF fuel pin design bases via the FFTF fuel pin design procedure.

Energy Technology Data Exchange (ETDEWEB)

An austenitic alloy having good thermal stability and resistance to sodium corrosion at 700.degree. C. consists essentially of 35-45% nickel 7.5-14% chromium 0.8-3.2% molybdenum 0.3-1.0% silicon 0.2-1.0% manganese 0-0.1% zirconium 2.0-3.5% titanium 1.0-2.0% aluminum 0.02-0.1% carbon 0-0.01% boron and the balance iron.

Energy Technology Data Exchange (ETDEWEB)

The US National Energy Policy of 2001 advocated the development of advanced fuel and fuel cycle technologies that are cleaner, more efficient, less waste-intensive, and more proliferation resistant. The need for advanced fuel development is emphasized in on-going DOE-supported programs, e.g., Global Nuclear Energy Initiative (GNEI), Advanced Fuel Cycle Initiative (AFCI), and GEN-IV Technology Development. The Directorates of Energy & Environment (E&E) and Chemistry & Material Sciences (C&MS) at Lawrence Livermore National Laboratory (LLNL) are interested in advanced fuel research and manufacturing using its multi-disciplinary capability and facilities to support a design concept of a small, secure, transportable, and autonomous reactor (SSTAR). The E&E and C&MS Directorates co-sponsored this Laboratory Directed Research & Development (LDRD) Project on Mono-Uranium Nitride Fuel Development for SSTAR and Space Applications. In fact, three out ...

International Nuclear Information System (INIS)

The RESF is utilized for storage of spent fuel under emergency conditions as well as for testing of FM heads. It receives cooling supply from the PHT Pressurizing pumps and after removal of decay heat from the spent fuel it goes to the D2O Storage Tank. The geometry of the RESF system is such that it can not sustain the thermosyphon loop during SBO, due to high frictional forces. To achieve the sustained thermosyphon, modifications in the design were suggested viz., removal of the steam trap and the relief valve above it and replacement by a solenoid valve (SV-16). In the event of SBO, SV-16 will open on 'RESF channel temperature high' signal and connect to FT D2O tank. The tank, being at atmospheric pressure and at lower elevation, will provide higher cooling flow rate through the RESF channel. D2O is periodically removed from the FT D2O tank by operating a Class-II pump intermittently. The analysis was done for the modified RESF system configuration to estimate the flow rate through ...

Science.gov (United States)

A high power AlGaInP single quantum well graded index separate confinement heterostructure. It comprises a substrate and a multiplicity of layers deposited thereon comprising a single Ga{sub x}In{sub x}P quantum well where x has a value from about 0.4 to about 0.6; multiple graded index regions on both sides of the quantum well and cladding layers adjacent to each graded region of the well, the graded region comprising Al{sub y}(Ga{sub 1{minus}y}){sub 0.5}In{sub 0.5}P quaternary alloy; wherein the value of y in the graded region varies from about 0.2 at the quantum well/graded region interface to up to about 0.6 for the cladding layers/graded index regions; the heterostructure having a low broad area threshold current with pulsed thresholds in the range from about 1 to about 2 Amps/cm{sup 2} and a differential efficiency of from about 20 to about 60 percent.

British Library Electronic Table of Contents (United Kingdom)

Laser consolidation (LC) developed by National Research Council???s Industrial Materials Institute (NRC-IMI-London) since mid-1990s, is a laser cladding based rapid manufacturing and material additive process that could fabricate a ???net-shape??? functional metallic shape through a ???layer-upon-layer??? deposition directly from a computer aided design model without using molds or dies. In order to evaluate the LC processability of different materials, some representative nickel-based superalloys (IN-625, IN-718, IN-738, and Waspaloy), stainless steels (austenitic SS316L and martensitic SS420), and lightweight alloys (Ti???6Al???4V titanium alloy and Al-4047 aluminum alloy) have been investigated. Like other laser cladding based processes, due to process-induced rapid directional solidifi...

Energy Technology Data Exchange (ETDEWEB)

JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on ...

Energy Technology Data Exchange (ETDEWEB)

Experience at SRS, ORNL, BNL, and Georgia Institute of Technology involving irradiated aluminum clad fuel and target elements, as well as studies of non-irradiated aluminum indicate that some types of aluminum assemblies can be kept in a continually well-deionized water atmosphere for up to 25 years without problems. SRS experience ranges from 2.75 years for the L-1.1 charge kept in deionized D{sub 2}O{sup 1} to greater than 10 years for assemblies stored in the Receiving Basin for Off-site Fuel (RBOF){sup 2}. Experience at Georgia Institute of Technology reactor in Atlanta yielded the longest value of 25 years without problems. The common denominators in all of the reports is that the water is continually deionized to approximately 2 M{Omega} (2 {times} 10{sup 6}ohms) resistivity and the containers for the water are stainless steel or other non-porous material. This resistivity value is equivalent to a value of 0.5 micromhos or microSiemens conductivity and is ...

Energy Technology Data Exchange (ETDEWEB)

We report tests of a refuelable zinc/air battery of modular, bipolar-cell design, intended for fleet electric busses and vans. The stack consists of twelve 250-cm{sup 2} cells built of two units: (1) a copper-clad glass-reinforced epoxy board supporting anode and cathode current collectors, and (2) polymer frame providing for air- and electrolyte distribution and zinc fuel storage. The stack was refueled in 4 min. by a hydraulic transfer of zinc particles entrained in solution flow.

Science.gov (United States)

This letter describes the first lasing experiment of the heterostructure semiconductor Raman laser with lateral confinement of both the Stokes and pump fields. It has a GaP Raman active layer with thickness of 10 ..mu..m and Al/sub 0.1/Ga/sub 0.9/P cladding layers. The stripe of the active layer has been fabricated by a plasma etching technique. Steps should be taken to realize the semiconductor Raman laser pumped by an injection laser, applicable to wideband optical communication.

International Nuclear Information System (INIS)

Fuel rod failures produce a release of fission products into primary coolant system. Since nuclear power plants have licensing limits for the release of volatile fission products to the environment (off-gas limits) detailed monitoring of the development of clad failure is necessary. In case of fuel rod failure a release of fission products into the primary coolant system arises. Fission gases accumulated in the free volume of a fuel rod escape through the clad defect. Water entering the fuel rod reacts with fission products, forming volatile chemical compounds. These may escape in a similar manner into the fission gases. Other compounds may dissolve and may be carried outside the fuel rod as dissolved species. Consequently, the distribution of these fission products, in the cross section of the fuel rod, is modified. An implementation of the maximum entropy gamma computed tomography technique is used to obtain such distributions in the area of ...

International Nuclear Information System (INIS)

The General Electric Test Reactor emergency cooling system performance was tested by intentionally scramming the reactor and then terminating the power to the primary pump. Certain transient thermal-hydraulic data were obtained preceding and during the established natural convection cooling loop composed of the upward flow through the core and the downward flow through the pool. An analysis was performed to permit the data to be extrapolated to obtain distributed fuel element flow rates and bulk temperature rises during the established cooling loop. The earliest time for the quasi-steady natural cooling loop to develop is about 2.5 min following scram. The cladding hot-spot temperature does not exceed the local saturation temperature after quasi-steady flow is established. Data are presented to assist in the modeling of the GETR natural convection loop. Semi-empirical relationships for friction factor and Nusselt number are also presented.

Energy Technology Data Exchange (ETDEWEB)

This paper summarizes the results of a long-term research project by SKBV which aimed to find ways of increasing spiral chute capacity while minimizing particle degradation, and also to extend the life of chute linings. Spiral chutes have been redesigned in such a way that chute speeds are considerably slower. The most wear-resistant and economic material for the inlet area was found to be slabs of tempered alloy cast steel, and for normal chute sections, cast basalt plates. Vertical rom bunkers can now be lined with thin cladding sections of GRC. Staple pit spiral chutes have been considerably improved in respect to wear and abrasion by the use of toughened steel as the chute lining material. Several of the innovations developed during the course of the research project are in use in West German mines and some have also been introduced abroad. (In German)

International Nuclear Information System (INIS)

Being based on the Segal method, calculation was carried out for the natural uranium nuclear fuel cost with Zircaloy-4 cladding having design parameters of Wolsung Nuclear Power Plant, CANDU-PHWR (Unit 1), currently under construction in Korea aiming at its completion in 1982. An attempt was also made for the sensitivity analysis of each fuel component; i.e., depreciation of fuel manufacturing plant caused by its life time, its load factor, production scale expansion of plant facilities, variations of construction and operating costs of fuel manufacturing plant, fluctuation of interest rates, extent of uranium ore price increases and effect of learning factor. (author).

International Nuclear Information System (INIS)

By varying different variables (potential, chloride content, surface pretreatment etc.) it is possible to show the effect of increasingly harsher corrosion conditions on passive-layer stability. These changes will not be detected by conventional methods e.g. measurement of passive current density, until a very late stage. Electrochemical noise measurement will allow the action of inhibitors to be detected in a rapid and effective manner. High-alloy steels and titanium claddings subjected to pitting corrosion in chloride-containing media demonstrate how to use this method for problem solutions and to open up new practical applications. The effect of mechanical stresses (natural, tensile) on the corrosion system can be detected by electrochemical noise which opens up new methods for crack corrosion studying and monitoring. (orig./DG).

Science.gov (United States)

Development of a shear and leach complex for UO/sub 2/-SS clad fuel was continued with major emphasis on the operation of a rotary drum leacher. Flooding data for nozzle plate pulse columns at high A/O ratios are reported. Engineering tests of dissolution of Zr--6% U alloy with HF in molten salt demonstrated a dissolution rate of 0.8 to 1.5 mg/cm/sup 2//min. Subsequent fluorination with 100% F/sub 2/ proceeded at half times of 40 to 135 min. The results of a high level waste calcination (run R-72) made with formaldehyde treated simulated Purex waste are reported. (auth)

International Nuclear Information System (INIS)

Zirconium alloys have anisotropic mechanical properties depending on their physical orientations and are widely used as nuclear materials such as cladding tube material. An operation condition of the nuclear reactor requires a high creep resistance, because it is subjected to long period operations, high temperature and high pressure. Generally, it takes a few days or months to do the creep experiment, so it is difficult to get a data in short period. However, there is a way to predict a creep property by using the stress-relaxation in the short term. These studies realized the stress-relaxation through a compressive test of HANA-6 (Zr-1.1Nb-0.05Cu) alloy that was developed by KAERI (Korea Atomic Energy Research Institute), and then predicted the creep property

International Nuclear Information System (INIS)

It is generally known that a number of metallic glasses have excellent corrosion resistance in a variety of chemically hostile environments. Consequently, the use of laser cladding to coat a massive crystalline material such as aluminium with a layer of a metallic glass has obvious advantages. In this paper, the authors will show that the formation of a predominantly amorphous layer of Zr_6_0Al_1_5NI_2_5 alloy by laser processing is possible, if the obstacles to amorphization are overcome. In addition, evidence of the excellent corrosion resistance of this alloy in a NaCl solution will be given. A comparative study of the corrosion behavior of this amorphous alloy with pure aluminium and Al-Cr alloy will be done, in order to complete previous studies of laser processed coatings of aluminum substrates.

International Nuclear Information System (INIS)

An effective procedure for removing _3H, Xe and Kr from irradiated fuels was demonstrated using Shippingport UO"2 fuel. The release characteristics of _3H, Kr, Xe, and I from irradiated nuclear fuel have been determined as a function of temperature and gaseous environment. Vacuum outgassing and a flowing gas stream have been used to vary the gaseous environment. Vacuum outgassing released about 99% of the _3H and 20% of both Kr and Xe within a 3 h at 1500_0C. Similar results were obtained using a carrier gas of He containing 6% H"2. However, a carrier gas containing only He resulted in the release of approximately 80% of the _3H and 99% of both Kr and Xe. These results indicate that the release of these volatile fission products from irradiated nuclear fuel is a function of the chemical composition of the gaseous environment. The rate of tritium release increased with increasing temperature (1100 to 1500_0C) and with the addition of hydrogen to the gas stream. Using crushed UO"2 fuel ...

International Nuclear Information System (INIS)

TRIGA-SSR is a steady state research and material test reactor that has been in operation since 1980. The original TRIGA fuel was HEU (highly enriched uranium) with a U"2"3"5 enrichment of 93 per cent. Almost all TRIGA HEU fuel bundles are now burned-up. Part of the spent fuel was loaded and transferred to US, in a Romania - DOE arrangement. The rest of the TRIGA fuel bundles have to be temporarily stored in the TRIGA facility. As the storage conditions had to be established with caution, neutron and thermal hydraulic evaluations of the storage conditions were required. Some criticality evaluations were made based on the SAR (Safety Analysis Report) data. Fuel constant axial temperature approximation effect is usual for criticality computations. TRIGA-SSR fuel bundle geometry and materials model for SCALE5-CSAS module allows the introduction of a fuel temperature dependency for the entire fuel active height, using different materials for each fuel bundle region. Previous RELAP5 thermal ...

Energy Technology Data Exchange (ETDEWEB)

The first of three scheduled poolside fuel examinations at the Oyster Creek reactor conducted during February/March 1980, was directed at one of the four symmetrically loaded ENC 8 x 8 lead assemblies that had achieved a burnup of approx. 25,000 MWd/MTU. Forty-five of the fuel rods in assembly UD3-109 were removed and examined. In general, the individual fuel rods were in excellent condition. The average fuel rod diameter continued to decrease during the last cycle was assembly burnup increased from 19,500 to 25,700 MWd/MTU. The creepdown since the beginning of life (BOL) in the center of the fuel rods is about 0.003 in. The fuel rods bore no indication of cladding ridging. Fuel rod growth continued at a linear rate of about 0.02% per GWd/MTU burnup since BOL. Preliminary eddy current test data showed that the cladding was free of significant defects. Visual examination of fuel rods, spacers, and tie plate revealed no unusual conditions. An ...

Energy Technology Data Exchange (ETDEWEB)

Plans for the disposal of radioactive liquid and solid wastes presently stored in double-shell tanks at the Hanford Site call for retrieval and processing of the waste to create forms suitable for permanent disposal. Waste will be retrieved from a tank using a submerged slurry pump in conjunction with one or more rotating slurry jet mixer pumps. Pacific Northwest Laboratory (PNL) has conducted tests using simulated waste slurries to assess the effects of a impinging slurry jet on the corrosion rate of the tank wall and floor, an action that could potentially compromise the tank`s structural integrity. Corrosion processes were investigated on a laboratory scale with a simulated neutralized cladding removal waste (NCRW) slurry and in a subsequent test with simulated neutralized current acid waste (NCAW) slurry. The test slurries simulated the actual NCRW and NCAW both chemically and physically. The tests simulated those conditions expected to exist in the respective ...

International Nuclear Information System (INIS)

Fourteen irradiated Fast Flux Test Facility (FFTF) fuel pins were subjected to representative overpower transients in six flowing sodium loop experiments conducted in the TREAT reactor. The transient tests were extended to substantial overpower levels well beyond protected levels, with some tests intentionally run to failure to identify failure thresholds and characteristics. Test variables included transient ramp rate (5, 50, and 100 cents/s) and burnup (2 to 58 MWd/kg). Performance limits and failure characteristics were identified, and cladding strain and fuel melting data were obtained for development and verification of transient analysis codes. The test results demonstrated that FFTF Reference fuel pins are capable of surviving overpower levels well beyond the FFTF secondary Plant Protection System (PPS) trip limit of 1.25 times normal rated power. Based on analytical evaluations to interpolate and extrapolate test results to the full range of burnups and ...

Energy Technology Data Exchange (ETDEWEB)

Continuous-wave (cw) operation at temperatures up to 23 /sup 0/C of an Al/sub 0.26/Ga/sub 0.26/In/sub 0.48/P/Ga/sub 0.52/In/sub 0.48/P/ Al/sub 0.26/Ga/sub 0.26/In/sub 0.48/P double heterostructure (DH) laser has been achieved for the first time. The threshold current was 160 mA at 20 /sup 0/C for a device with a 10-..mu..m-wide and 250-..mu..m-long ion-implanted stripe geometry. The emission wavelength was 671 nm during cw operation at 10 /sup 0/C. To reduce thermal resistance to a heat sink, a dually stacked structure made of a thin (approx.0.3 ..mu..m) p-AlGaInP layer and a p-Al/sub 0.76/Ga/sub 0.24/As layer was used as a cladding layer. The DH wafer was grown by atmospheric pressure metalorganic chemical vapor deposition.

International Nuclear Information System (INIS)

In the VEGA program on radionuclide release from irradiated fuel under severe accident conditions, 10 tests in total were performed at JAEA from 1999 to 2004 under inert and steam atmospheres including the highest pressure or temperature conditions. These tests showed the increase in release rate above 2,800 K or at the fuel liquefaction and the decrease in release rate under elevated pressure, which was a first observation in the world. The data on low-volatility radionuclide release, release from MOX fuel, effect of fuel oxidation, and eutectic reaction with cladding on release were obtained from the tests. The mechanism of pressure effect on release was examined and a new release model with pressure effect was proposed. In addition, the pressure effect on source term evaluation and effectiveness of accident management measures were investigated. This article summarizes the major outcomes described above that have already been published and newly describes the ...

International Nuclear Information System (INIS)

The Sp-100 reactor is a lithium-cooled high-temperature fast-spectrum reactor. The fuel is UN. The cladding is fabricated from PWC-11, a Nb alloy, as are all the primary structural components. A reactor lifetime of up to ten years with an operating temperature of 1370 K is required. The accumulated fluence is expected to be 6 x10"2"2 n/cm"2. The damage, which could result in swelling or embrittlement, anneals out as fast as it occurs for the majority of the structure. This has been confirmed by earlier radiation testing. A number of components, however, are exposed to lower temperatures and the reactor design and materials selection for these components must take this into consideration. Radiation effects must also be considered for the UN fuel, bearing materials, etc. To data an instrumented experiment, MOTO 1000A, has been conducted in the FFTF reactor and as uninstrumented experiment SPM-1 in the EBR-II reactor. In this paper the post-irradiation examination and ...

Energy Technology Data Exchange (ETDEWEB)

This paper presents the new thermal hydraulic models describing the hydrodynamics of the solid fuel/steel chunks during an LMFBR hypothetical core-disruptive accident. These models, which account for two-way coupling between the solid and fluid phases, describe the mass, momentum, and energy exchanges which occur when the chunks are present at any axial location. They have been incorporated in LEVITATE, a code for the analysis of fuel and cladding dynamics under Loss-of-Flow (LOF) conditions. Their influence on fuel motion is presented in the context of the L6 TREAT experiment analysis. It is shown that the overall hydrodynamic behavior of the molten fuel and solid-fuel chunks is dependent on both the size of the chunks and the power level. At low and intermediate power levels the fuel motion is more dispersive when small chunks, rather than large ones, are present. At high power levels the situation is reversed. These effects are explained in detail.

Energy Technology Data Exchange (ETDEWEB)

The electrometallurgical treatment process described in this paper builds on our experience in treating spent fuel from the Experimental Breeder Reactor (EBR-II). The work is also to some degree, a spin-off from applying electrometallurgical treatment to spent fuel from the Hanford single pass reactors (SPRs) and fuel and flush salt from the Molten Salt Reactor Experiment (MSRE) in treating EBR-II fuel, we recover the actinides from a uranium-zirconium fuel by electrorefining the uranium out of the chopped fuel. With SPR fuel, uranium is electrorefined out of the aluminum cladding. Both of these processes are conducted in a LiCl-KCl molten-salt electrolyte. In the case of the MSRE, which used a fluoride salt-based fuel, uranium in this salt is recovered through a series of electrochemical reductions. Recovering high-purity uranium from an aluminum-matrix fuel is more challenging than treating SPR or EBR-II fuel because the aluminum- matrix fuel is typically -90% ...

Energy Technology Data Exchange (ETDEWEB)

Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility and cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing ...

International Nuclear Information System (INIS)

A suspension-firing type radioactive waste incineration system is developed and cold demonstration testing of ceramic filters for the system are carried out. The incineration system, which is useful for a wide variety of waste materials, can serve to simplify the facilities and to reduce the costs for waste disposal. The incineration system can be used for drying-processing of concentrated waste liquids and disposal of flame resistant materials including ion exchange resins and rubber, as well as for ordinary combustible solid materials. An on-line backwash system is adopted to allow the ceramic filters to operate stably for a long period of time. For one-step filtering using the ceramic filter, the decontamination factor is greater than 10"5 for the processing of various wastes. In a practical situation, there exist vapor produced by the spray drier and the cladding in used ion exchange resin, which act to increase the decontamination performance of the ceramic ...

International Nuclear Information System (INIS)

During 17 years since the start of operation of the first commercial LWR in Japan, many LWRs have experienced various corrosion damages, but the causes of them were clarified, and the counter-measures were executed effectively in actual plants, as the results, the cause of corrosion damage decreased remarkably, and now, the high rate of operation has become to be maintained. In this paper, the major cases of corrosion damage experienced in LWRs in Japan and foreign countries, the causes of them and the countermeasures, the problems of hereafter and so on are described. The corrosion damage of metallic materials in the environment of LWRs occurs in the parts in contact with high temperature, high pressure water and steam, such as stainless steel piping in the primary cooling system of BWRs, and nickel alloy heating tubes of steam generators, carbon steel feed water piping and zirconium alloy fuel cladding tubes in PWRs. Recently, accompanying the advance of LWR ...

International Nuclear Information System (INIS)

Further to a cost-benefit analysis of the various medium-term and long-term and H.L.W. storage possibilities, C.E.A. (French Atomic Energy Commission) and S.G.N. decided to develop an original dry storage process with natural convection cooling that offers many advantages: cut in the total investment and operating costs; high operating safety; natural convection cooling; existence of two containment barriers irrespective of the assumed clad conditions; flexible, modular and compact design. The process was first implemented in the so-called CASCAD Cadarache Facility (vault-type facility) constructed in Cadarache mainly to store fuel from Brennilis heavy water reactor. For the purpose, a large program was set up to develop and validate computer codes, in particular with the use of mockups. On the request of many clients, and owing to the outstanding operating results of the CASCAD Cadarache Facility, SGN was brought to adapt the process to the storage of other types ...

International Nuclear Information System (INIS)

The TRAC/PFI-MOD2 Version 5.3 code was assessed against six FLECHT-SEASET forced reflood tests (31504, 31203, 31302, 31701, 34209, and 31922) and two cylindrical core test facility (CCTF) tests [C1-19 and C2-6]. The objective of this study was to evaluate the clad thermal response predictive capabilities of the code with the newly added reflood model under large-break loss-of-coolant accident (LOCA) conditions in a pressurized water reactor (PWR). The TRAC model for the FLECHT-SEASET test facility was developed from a RELAP5 model. The test section was modeled using a vessel component with 23 axial levels, 1 radial ring, and 1 azimuthal cell. Test inlet and exit conditions were modeled using fill and break components, respectively. The measured lower and upper plenum test conditions were input to the model. The electrically heated rod was modeled using a rod component with 22 axial mesh points. The axial boundary of each mesh point coincided with a fluid cell ...

Science.gov (United States)

Stresses commonly present in AlGaAs/GaAs laser heterostructures were reduced using Molecular Beam Epitaxy grown Al/sub 0.3/Ga/sub 0.7/P/sub 0.01/As/sub 0.99/ cladding layers. The Al/sub 0.3/Ga/sub 0.7/P/sub 0.01/As/sub 0.99/ alloy was grown using an incident P/sub 2/ flux of roughly-equal1 x 10/sup 14//cm/sup 2/ indicating a sticking coefficient of 0.1 at a substrate temperature of 600 /sup 0/C. X-ray automatic Bragg angle control curvature measurements were used to monitor the residual heteroepitaxial stress. Broadened double crystal x-ray linewidths indicated the occurrence of alloy grading and broadened interfaces. The effects of P concentration and film thickness on stress and on the existence of a misfit dislocation grid are discussed.

Energy Technology Data Exchange (ETDEWEB)

Production of cermet nuclear fuel having fine uranium dioxide (UO{sub 2}) particles dispersed in matrix metal requires basic property data on the compatibility of matrix metal with fission product compounds. It is thermodynamically suggested that, as burnup increases, cesium in oxide fuel reacts with the fuel, other fission products or cladding pipe and produces cesium uranates, cesium molybdate, or cesium chromate in stainless steel cladding pipe. Attempt was made to measure the thermal expansion coefficient and thermal conductivity of cesium uranates (Cs{sub 2}UO{sub 4} and Cs{sub 2}U{sub 2}O{sub 7}), cesium molybdate (Cs{sub 2}MoO{sub 4}) and cesium chromate (Cs{sub 2}CrO{sub 4}). Thermal expansion was measured by X-ray diffraction and determined by Cohen`s method. Thermal conductivity was obtained by measuring thermal diffusion by laser flash method. The thermal expansion of Cs{sub 2}UO{sub 4} and Cs{sub 2}U{sub 2}O{sub 7} is as low as 1.2% ...

Energy Technology Data Exchange (ETDEWEB)

Pipeline automated ultrasonic testing (AUT) uses fully automated equipment that travels around a pipe on a welding band in a linear scan to allow array pulses to provide ultrasound information on weld zones. Pipeline AUT is also used for zone discrimination as well as with special calibration blocks, dual gate output displays, and rapid defect sizing. AUT allows for the inspection of welds soon after completion, and can save construction costs by process control when combined with engineering critical assessment (ECA) to minimize rejection rates. Ultrasonic phased arrays (UPA) use an array of elements that are all individually wired, pulsed, and time-shifted. Elements are pulsed in groups of approximately 16 elements at a time for pipeline welds. Phased arrays can be used to combine electronic scanning, sectorial scanning and precision focusing to give a practical combination of displays. Optimum angles can be selected for welds and other components, and the electronic scanning ...

International Nuclear Information System (INIS)

This study deals with testing the capacity of the code PARET to simulate natural circulation phenomena under different boundary conditions in addition to assessment of some new options related to simulation of control rod movement and the reactivity effect of thermal expansion fuel elements. the experiments of the simple thermal hydraulic loop of Missouri University about natural circulation phenomena in narrow parallel channel were used to validate the code. The results indicate good agreements regarding the evolution of coolant velocity and clad temperature. In particular the heat transfer coefficient of natural convection has been calculated in good agreement with the experiment. On the other hand, the core of MNSR reactor has been modelled to stimulate the reactor dynamic behaviour under natural circulation condition for different initial power level. The observed oscillations during the initial phase vanish gradually with passing time. In this context three ...

International Nuclear Information System (INIS)

Important steady-state thermohydraulic parameters of the TRIGA research reactor operating under natural convection mode of coolant flow were investigated using NCTRIGA computer code. Neutronic parameters used in preparing the input of NCTRIGA were taken from the analysis performed by 3-D Monte Carlo code MCNP4C. Benchmarking of the NCTRIGA calculated results were performed against the experimental data measured by the thermocouples in the instrumented fuel element (IFE) during the steady state operation of the reactor under natural convection mode of coolant flow. Various thermohydraulic parameters like the coolant velocity, flow rate and mass flow rate were generated for the hot channel as well as for the two channels comprising instrumented fuels. Calculated peak fuel temperatures at different power levels were compared with the measured values and also with the calculations performed by PARET code. Axial temperature profile at the fuel centreline, fuel surface and coolant in the hot ...

Science.gov (United States)

We present results from the first studies of electric-field effects on optical transitions in visible-band-gap InGaP/InAlGaP multiple-quantum-well (MQW) structures. These structures, grown at 775 [degree]C by metalorganic vapor phase epitaxy on (100) GaAs substrates misoriented 6[degree] towards P(111)[r angle][l angle]111[r angle]A, consist of nominally undoped MQWs surrounded by doped In[sub 0.49]Al[sub 0.51]P cladding layers to form [ital p]-[ital i]-[ital n] diodes. The Stark shifts of various allowed and forbidden quantum-well transitions were observed in bias-dependent electroreflectance spectra of In[sub 0.49]Ga[sub 0.51]P/In[sub 0.49](Al[sub 0.5]Ga[sub 0.5])[sub 0.51]P MQW samples with 10-nm-thick layers. We find the magnitude of these shifts to depend on the details of the Mg doping profile, confirming the importance of Mg diffusion and unintentional background doping in these materials. Our results show that (InAlGa)P materials are promising for ...

International Nuclear Information System (INIS)

In Japan, the development of next-generation Light Water Reactor has been launched since April 2008. The development program will be completed in 2015. The purpose of development is to cope with the replacement for existing nuclear power plants after 2030 in Japan and the expanding demand for nuclear power in the world; 'Nuclear Renaissance.' The reactor also aims to be global standard at around 2030. The requirements for global standard and domestic users have been investigated through the feasibility study of past 2 years, 2006-2007, and six innovative features or 'Core-Concepts' were established as follows. A) Reactor core system with uranium enrichment above 5% for significant decrease of spent fuel discharge and prominent higher availability B) Long-life materials and innovative water chemistry technologies for 80 years plant lifetime and significant reduction of occupational dose C) Seismic isolation technologies to standardize plant design independent from site conditions D) ...

Energy Technology Data Exchange (ETDEWEB)

The Fixed Bed Nuclear Reactor (FBNR) is a small 40 MWe reactor based on the Pressurized Water Reactor (PWR) technology. FBNR is an integrated primary circuit and simple in design. It has the characteristics of being small, modular, proliferation resistant, inherently safe and passively cooled reactor with reduced adverse environmental impact. It utilizes the fuel designed for high temperature reactors operating in a relatively low temperature of PWR environment The 15 mm diameter spherical fuel elements are made of TRISO type microspheres embedded in graphite and cladded by SiC. The coolant flow transfers them from the fuel chamber into the core and become fixed forming a suspended core. Any accident signal will cut off the power to the coolant pump causing a stop in the flow. This results in making the fuel elements fall out of the reactor core by the force of gravity and return into the fuel chamber where they are passively cooled under subcritical condition. ...

Energy Technology Data Exchange (ETDEWEB)

The COBRA-SFS computer code was used to perform thermal-hydraulic analyses of consolidated spent fuel stored in casks. The ability of the COBRA-SFS code to model consolidated fuel was evaluated by comparing predictions with experimental data obtained from electrically heated rod bundles by Ridihalgh, Eggers, and Associates and Eggers Ridihalgh Partners, Inc. under sponsorship of the Electric Power Research Institute. The calculations agreed with the measured temperatures well within the bounds of experimental error. Based on the evaluation results, best-estimate temperature predictions were performed for consolidated fuel in several cask designs. Results are presented for the REA 2023 BWR cask, the CASTOR-1C BWR cask, and the Concrete Sealed Storage Cask designed for Monitored Retrievable Storage (MRS). The cask simulation results indicate that consolidation of spent fuel results in a reduction of convection and radiation heat transfer from the fuel rods while enhancing conduction heat ...

Energy Technology Data Exchange (ETDEWEB)

The Global Nuclear Energy Partnership (GNEP) is proposing to develop a sodium-cooled fast-spectrum reactor (SFR) to transmute and consume actinides from discharged nuclear fuel. To meet performance objectives, new and advanced fuels and targets need to be developed. The fuels to be irradiated include metal and oxide mixed actinides (U-Np-Pu-Am-Cm); for the target concept, Am-Cm has been considered. A significant part of the development process is the irradiation of the fuel and cladding in a prototypic fast reactor environment to determine the performance under irradiation. Analysis results are presented in this paper for a fast-neutron irradiation facility design based on the large fast neutron flux available in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) combined with the use of a strongly-absorbing thermal neutron shield. Several designs were assessed; the preferred concept consists of a three-pin design with an europium oxide ...

Energy Technology Data Exchange (ETDEWEB)

Micron-size Ni-base alloy (NBA) powders were mixed with both 1.5 wt.% (hereinafter %) micron-size CeO{sub 2} (m-CeO{sub 2}) and also 1.5% and 3.0% nano-size CeO{sub 2} (n- CeO{sub 2}) powders. These mixtures were coated on low-carbon steel (Q235) by 2.0 kW CO{sub 2} laser cladding. The effects on the microstructures, phases and electrochemical corrosion of the coatings upon the addition of m- and n- CeO{sub 2} powders to NBA (m- and n- CeO{sub 2} /NBA) have been investigated. The results showed that a smooth coating was prepared under suitable processing parameters (P= 2.0 kW, V= 180 mm min{sup -1}) by adding 1.5% n- CeO{sub 2}. In addition to the primary phases of {gamma}-Ni, Cr{sub 23} C{sub 6} and Ni{sub 3} B in the Ni-base alloy coating, CeNi{sub 3} was formed in Ni-base alloy coatings with both n- CeO{sub 2} and m-CeO{sub 2} particles, and CeNi{sub 5} appeared in the coating upon decreasing the size of CeO{sub 2} particles. Well-developed dendrites were ...

International Nuclear Information System (INIS)

Nuclear power plants of 20 units of in Korea are generating about 700 MTU of spent fuels annually. The inventory of spent fuels in Korea were estimated about 10,087.07 MTU at end of 2008, and the storage space of spent fuels won't be available any more at 2016 due to the saturation of the spent fuel pools in the plants. In addition, in order to reduce carbon emission and correspond to the enormous electricity demand in Korea, 8 units of nuclear power plants are under construction and several more plants are under planning. The 100,000 MTU of spent fuel inventory are expected by the year of 2095 in Korea. Therefore, short term and long term of spent fuel management plans are under discussion and implementation in Korea. As a short term of spent fuel management strategy for the target year of 2016, central or local spent fuel dry interim storage options are mostly under discussion. As a long term of management plan, fast reactor and advanced fuel cycle R and D plan were approved by 255th ...