Post-irradiation examination of Al-61 wt% U3Si fuel rods from the NRU reactor

International Nuclear Information System (INIS)

Sears, D.F.; Wang, N.

1997-01-01

This paper describes the post-irradiation examination of 4 intact low enrichment uranium (LEU) fuel rods from the national research universal (NRU) reactor at the Chalk River Laboratories of AECL. The rods were irradiated during the period 1993 through 1995, under typical driver fuel operating conditions in NRU, i.e., nominal D 2 O coolant inlet temperature 37E C, inlet pressure 654 kPa and mass flow 12.4 L/s. Irradiation exposures ranged from 147 to 251 full-power days, corresponding to 40 to 84 atom % 235 U burnup. The maximum rod power was ∼2 MW, with element linear power ratings up to 68 kW/m. Post-irradiation examinations, conducted in 1997, focused on optical metallography to measure cladding oxide thickness and fuel core and cladding microstructural examinations. The cladding oxide was approximately 24 : m thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 : m thick on the fins. The cladding retained significant ductility after irradiation, and its microstructure appeared unchanged. Fuel core diametral increases were small (up to 4%) and within the range previously observed on Al-61 wt% U 3 Si fuel irradiated in the NRU reactor. (author)

Post-irradiation examination of fuel elements of Tarapur Atomic Power Station (Report-I)

International Nuclear Information System (INIS)

Bahl, J.K.; Sah, D.N.; Chatterjee, S.; Sivaramkrishnan, K.S.

1979-01-01

Detailed post-irradiation examination of three initial load fuel elements of the Tarapur Atomic Power Station (TAPS) has been carried out. The causes of the element failures have been analysed. It was observed that almost 90% of the length of the elements exoerienced nodular corrosion. It has been estimated that nodular corrosion would seriously affect the wall thickness and surface temperature of higher rated elements. Lunar shaped fret marks have also been observed at some spacer grid locations in the elements. The depth of the largest fret mark was measured to be 16.9% clad wall thickness. Detailed metallographic examination of the clad and fuel in the three elements has been done. The temperatures at different structural regions of the fuel cross-sections have been estimated. The change in fuel density during irradiation has been evaluated by comparing the irradiated fuel diameter with the mean pellet design diameter. The performance of the end plug welds and spacer grid sites in the elements has been assessed. The burnup distribution along the length of the elements has been evaluated by gamma scanning. The redistribution of fission products in the fuel has been examined by gamma scanning and beta-gamma autoradiography. Mechanical properties of the irradiated cladding have been examined by ring tensile testing. (auth.)

Post-irradiation examination of A1-61 wt % U3Si fuel rods from the NRU reactor

International Nuclear Information System (INIS)

Sears, D.F.; Wang, N.

1997-09-01

This paper describes the post-irradiation examination of 4 intact low-enrichment uranium (LEU) fuel rods from the national research universal (NRU) reactor at the Chalk River Laboratories of AECL. The rods were irradiated during the period 1993 through 1995, under typical driver fuel operating conditions in NRU, i.e., nominal D 2 0 coolant inlet temperature 37 degrees C, inlet pressure 654 kPa and mass flow 12.4 L/s. Irradiation exposures ranged from 147 to 251 full-power days, corresponding to 40 to 84 atom % 235 U burnup. The maximum rod power was ∼2 MW, with element linear power ratings up to 68 kW/m. Post-irradiation examinations, conducted in 1997, focused on optical metallography to measure cladding oxide thickness and fuel core and cladding microstructural examinations. The cladding oxide was approximately 24 μm thick at the mid-plane of fuel rods irradiated to 251 full-power days, with small areas up to 34 μm thick on the fins. The cladding retained significant ductility after irradiation, and its microstructure appeared unchanged. Fuel core diametral increases were small (up to 4%) and within the range previously observed on A1-61 wt % U 3 Si fuel irradiated in the NRU reactor. (author)

Post-irradiation examination of CANDU fuel bundles fuelled with (Th, Pu)O2

International Nuclear Information System (INIS)

Karam, M.; Dimayuga, F.C.; Montin, J.

2010-01-01

AECL has extensive experience with thoria-based fuel irradiations as part of an ongoing R&D program on thorium within the Advanced Fuel Cycles Program. The BDL-422 experiment was one component of the thorium program that involved the fabrication and irradiation testing of six Bruce-type bundles fuelled with (Th, Pu)O 2 pellets. The fuel was manufactured in the Recycle Fuel Fabrication Laboratories (RFFL) at Chalk River allowing AECL to gain valuable experience in fabrication and handling of thoria fuel. The fuel pellets contained 86.05 wt.% Th and 1.53 wt.% Pu in (Th, Pu)O 2 . The objectives of the BDL-422 experiment were to demonstrate the ability of 37-element geometry (Th, Pu)O 2 fuel bundles to operate to high burnups up to 1000 MWh/kgHE (42 MWd/kgHE), and to examine the (Th, Pu)O 2 fuel performance. This paper describes the post-irradiation examination (PIE) results of BDL-422 fuel bundles irradiated to burnups up to 856 MWh/kgHE (36 MWd/kgHE), with power ratings ranging from 52 to 67 kW/m. PIE results for the high burnup bundles (>1000 MWh/kgHE) are being analyzed and will be reported at a later date. The (Th, Pu)O 2 fuel performance characteristics were superior to UO 2 fuel irradiated under similar conditions. Minimal grain growth was observed and was accompanied by benign fission gas release and sheath strain. Other fuel performance parameters, such as sheath oxidation and hydrogen distribution, are also discussed. (author)

Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

International Nuclear Information System (INIS)

Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

1982-01-01

The document reports in detail the fuel pin fabrication data and describes the irradiation conditions and history. All the relevant results of the non-destructive and destructive post-irradiation examinations are reported. They include: visual inspection and chemical analysis of crud; length and diameter measurements; neutron radiography and gamma scanning; juncture tests and fission gas analysis (including residual gas in fuel samples); microscopy and alpha + beta/gamma autoradiography; microprobe investigations; burn-up and isotopic analysis; and hydrogen analysis in clad. The data and observations obtained are discussed in detail and conclusions are given. The irradiation and post-irradiation examinations of the R-109 pins have shown the safe, pre-calculable performance of LWR fuel pins containing mixed-oxide sphere-pac fuel with the fissile material mainly present in the large spheres

WWER fuel: Results of post irradiation examination

International Nuclear Information System (INIS)

Markov, D.V.; Smirnov, V.P.; Smirnov, A.V.; Polenok, V.S.; Perepelkin, S.O.; Ivashchenko, A.A.

2006-01-01

Experience in the field of fabrication, operation, testing and post-irradiation examinations (PIE) made it possible to settle the following requirements for a new generation of WWER nuclear fuel: - For WWER-1000 FA, the service life is no less than 5 years, 3 alternative fuel cycles (FC): 12 months x 4 FCs, 12 months x 5 FCs and 18 months x 3 FCs; - For WWER-440 FA, fuel cycle is 12 months x 5 FCs and a part of operating assembly is left for the 6. year; - High fuel burnup - up to 70 MWd/kgU; - Dimensional stability of FA and its components; - FA repairability; - Adaptability of fuel cycles; - Maintenance of maneuvering operating conditions at the NPP; - Reliability of control rod operation; - High serviceability level - FE leakage is no worse than 10-5 l/year. In order to provide the fulfillment of the above-given requirements, designers and production engineers have worked out cumulative measures and engineering solutions, which are introduced in development of a new generation fuel. Currently old design FA-M assemblies provided with steel skeleton are being operated in WWER-1000 reactors at Ukrainian and Bulgarian NPPs. As for Russian NPPs, new-type FAs are operated. These are advanced FAs (AFA), FA-A and FA-2 provided with zirconium alloy skeletons. A design of the second generation of WWER-440 operating assemblies was developed with respect to changes in some geometrical parameters, fastening of FEs in the lower grid (splinting was substituted for collet), usage of reinforcing rib under the lower grid, anti-debris filter and hafnium elements of junction unit as well as hafnium content decrease from 0.05 % mass down to 0.01% mass in zirconium materials. They are basic designs of FAs in order to be introduced in a five-year fuel cycle of WWER-440 NPPs in Czech Republic and Slovakia since 2005 and have got prospects for development. The operating experience of dismountable operating assemblies at the Loviisa NPP, vibration-proof operating assemblies at the

Some results on development, irradiation and post-irradiation examinations of fuels for fast reactor-actinide burner (MOX and inert matrix fuel)

International Nuclear Information System (INIS)

Poplavsky, V.; Zabudko, L.; Moseev, L.; Rogozkin, B.; Kurina, I.

1996-01-01

Studies performed have shown principal feasibility of the BN-600 and BN-800 cores to achieve high efficiency of Pu burning when MOX fuel with Pu content up to 45% is used. Valuable experience on irradiation behaviour of oxide fuel with high Pu content (100%) was gained as a result of operation of two BR-10 core loadings where the maximum burnup 14 at.% was reached. Post-irradiation examination (PIE) allowed to reveal some specific features of the fuel with high plutonium content. Principal irradiation and PIE results are presented in the paper. Use of new fuel without U-238 provides the maximum burning capability as in this case the conversion ratio is reduced to zero. Technological investigations of inert matrix fuels have been continued now. Zirconium carbide, zirconium nitride, magnesium oxide and other matrix materials are under consideration. Inert matrices selection criteria are discussed in the paper. Results of technological study, of irradiation in the BOR-60 reactor and PIE results of some inert matrix fuels are summarized in this report. (author). 2 refs, 1 fig., 3 tabs

16-rod-bundle: Irradiation in the MZFR and post-irradiation examinations

International Nuclear Information System (INIS)

Manzel, R.

1979-04-01

In the course of the irradiation of a 16-rod prototype bundle, the basis has been established for the irradiation of experimental fuel assemblies containing full-length PWR fuel rods in standard positions of the MZFR. The prototype bundle was discharged after an irradiation time of 284 full power days and a burnup of 11400 MWd/tU. The overall performance of the prototype bundle was highly satisfactory. Detailed post-irradiation examinations confirmed the good conditions of bundle structures and fuel rods. (orig.) [de

ORNL capability to conduct post irradiation examination of full-length commercial nuclear fuel rods

International Nuclear Information System (INIS)

Spellman, Donald J.

2007-01-01

Hot cells at Oak Ridge National Laboratory (ORNL) are nearing completion of a multi-year upgrade program to implement 21. century capabilities to meet the examination demands for higher burnup fuels and the future demands that will come from fuel recycling programs. Fuel reliability and zero tolerance for fuel failure is more than an industry goal. Fuel reliability is becoming a requirement that supports the renaissance of nuclear power generation. Thus, fuel development and management of new forms of waste that will come from programs such as the Global Nuclear Energy Partnership (GNEP) will require extensive use of the flexible, high-quality, technically advanced hot cells at ORNL. ORNL has the capability to perform post irradiation examination (PIE) of irradiated commercial nuclear fuel rods and the management structure to ensure a timely, cost-effective result. ORNL can: 1) Handle the transportation issues, 2) Perform macroscopic fuel rod examinations, 3) Perform microscopic fuel and clad examinations, and 4) Manage legacy material and waste disposal issues from PIE activities. All four of these items will be managed in a way that allows the customer day-to-day access to the results and data. Hot cell examination equipment that is necessary to determine the characteristics and performance of irradiated materials must operate in a hostile environment and is subject to long-term degradation that may result in reliability and quality assurance (QA) issues. ORNL has modernized its hot cell nuclear fuel examination equipment, installing state-of-the-art automated examination equipment and data gathering capabilities. ORNL is planning a major commitment to nuclear fuel examination and development, and future improvements will continue to be made over the next few years. (author)

Main results of post-irradiation examinations of new-generation fuel assemblies VVER-1000

International Nuclear Information System (INIS)

Zvir, E.; Markov, D.; Polenok, V.; Zhitelev, V.; Kobylyansky, G.

2009-01-01

To increase the competitiveness of Russian nuclear fuel at the foreign market and to improve its technical and economic performance in order to provide a necessary level of safety, it is necessary to solve certain important tasks: Increase of fuel burn-up; Extension of operational lifetime of fuel assemblies and operational reliability of nuclear fuel; Introduction of cost-beneficial and flexible fuel cycles. Alternative fuel assemblies TVSA VVER-1000 and TVS-2 are used as a basis to optimize the nuclear fuel and develop advanced fuel cycles for nuclear power plants with VVER-1000 reactor types. Four fuel assemblies TVSA operated during 1 and up to 6 reactor cycles, reference fuel assembly TVS-2 operated during three reactor cycles and achieved an average fuel burnup of 48MW·day/kgU as well as failed fuel assembly TVS-2 operated during one cycle were examined at RIAR in recent years. The main objectives of these examinations were to obtain experimental data in support of operational integrity of products or to find out reasons of their failure. The performed post-irradiation examinations confirmed the operational integrity of alternative fuel assemblies TVSA including their geometrical stability up to the average fuel burnup of 55 MW·day/kgU over the fuel assembly (FA) (up to the maximal fuel burnup of ∼73 MW·day/kgU in fuel rods) and of TVS-2 up to the average fuel burnup of 48 MW·day/kgU over the fuel assembly. The changes introduced in the design of VVER-1000 fuel assembly during the development of alternative fuel assembly TVSA and TVS-2 did not make any negative effect on fuel rods. It was proved that causes of fuel rod failure were not related to design features of fuel assemblies. The design features and operating conditions of fuel assemblies under examinations are briefly described. Post-irradiation examinations proved the geometrical stability of fuel assemblies TVSA and TVS-2 under operation up to the fuel burnup of ∼50 MW day/kgU, as for the

The post irradiation examination of three fuel rods from the IFA 429 experiment irradiated in the Halden Reactor

International Nuclear Information System (INIS)

Williams, J.

1979-11-01

A series of fuel rod irradiation experiments were performed in the Halden Heavy Boiling Water Reactor in Norway. These were designed to provide a range of fuel property data as a function of burn-up. One of these experiments was the IFA-429. This was designed to study the absorption of helium filling gas by the UO 2 fuel pellets, steady state and transient fission gas release and fuel thermal behaviour to high burn-up. This data was to be obtained as a function of fuel density, fuel grain size, initial fuel/cladding gap, average linear heat rating, burn-up and overpower transients. All the fuel is in the form of pressed and sintered UO 2 pellets enriched to 13 weight percent 235 U. All the rods were clad in Zircaloy 4 tube. The details of the experiment are given. The post irradiation examination included: visual examination, neutron radiography, dimensional measurements, gamma scanning, measurement of gases in fuel rods and internal free volume, burn-up analysis, metallographic examination, measurement of retained gas in UO 2 pellets, measurement of bulk density of UO 2 . The results are given and discussed. (U.K.)

Post-irradiation examination of prototype Al-64 wt% U3Si2 fuel rods from NRU

International Nuclear Information System (INIS)

Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D.

1997-01-01

Three prototype fuel rods containing Al-64 wt% U 3 Si 2 (3.15 gU/cm 3 ) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U 3 Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U 3 Si 2 powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37 degrees C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U 3 Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL's research reactors

Post-irradiation examination of a failed PHWR fuel bundle of KAPS-2

International Nuclear Information System (INIS)

Mishra, Prerna; Unnikrishnan, K.; Viswanathan, U.K.; Shriwastaw, R.S.; Singh, J.L.; Ouseph, P.M.; Alur, V.D.; Singh, H.N.; Anantharaman, S.; Sah, D.N.

2006-08-01

Detailed post irradiation examination was carried out on a PHWR fuel bundle irradiated at Kakrapar Atomic Power Station unit 2 (KAPS-2). The fuel bundle had failed early in life at a low burnup of 387 MWd/T. Non destructive and destructive examination was carried out to identify the cause of fuel failure. Visual examination and leak testing indicated failure in two fuel pins of the outer ring of the bundle in the form of axial cracks near the end plug location. Ultrasonic testing of the end cap weld indicated presence of lack of fusion type defect in the two fuel pins. No defect was found in other fuel pins of the bundle. Metallographic examination of fuel sections taken from the crack location in the failed fuel pin showed extensive restructuring of fuel. The centre temperature of the fuel had exceeded 1700 degC at this location in the failed fuel pin, whereas fuel centre temperature in the un-failed fuel pin was only about 1300 degC. Severe fuel clad interaction was observed in the failed fuel pin at and near the location of failure but no such interaction was observed in the un-failed fuel pins. Several incipient cracks originating from the inside surface were found in the cladding near failure location in addition to the main through wall crack. The incipient cracks were filled with interaction products and hydride platelets were present at tip of the cracks. It was concluded from the observations that the primary cause of failure was the presence of a part-wall defect in the end cap weld of the fuel pins. These defects opened up during reactor operation leading to steam ingress into the fuel, which caused high fuel centre temperature and severe fuel-cladding interaction resulting in secondary failures. A more stringent inspection and quality control of end plug weld during fabrication using ultrasonic test has been recommended to avoid such failure. (author)

State of the VVER-1000 spent U-Gd fuel rods based on the results of post-irradiation examinations

International Nuclear Information System (INIS)

Shevlyakov, G.; Zvir, E.; Strozhuk, A.; Polenok, V.; Sidorenko, O.; Volkova, I.; Nikitin, O.

2015-01-01

The present paper is devoted to post-irradiation examinations (PIE) of U-Gd fuel rods with different geometry of the fuel pellets irradiated as part of the VVER-1000 fuel assembly. As evidenced by their PIE data, they did not exhaust their service life based on the main parameters (geometrical dimensions, corrosion state, and release of fission product gases). (author)

Hot cell facilities for post irradiation examination

International Nuclear Information System (INIS)

Mishra, Prerna; Bhandekar, Anil; Pandit, K.M.; Dhotre, M.P.; Rath, B.N.; Nagaraju, P.; Dubey, J.S.; Mallik, G.K.; Singh, J.L.

2017-01-01

Reliable performance of nuclear fuels and critical core components has a large bearing on the economics of nuclear power and radiation safety of plant operating personnel. In view of this, Post Irradiation Examination (PIE) is periodically carried out on fuels and components to generate feedback information which is used by the designers, fabricators and the reactor operators to bring about changes for improved performance of the fuel and components. Examination of the fuel bundles has to be carried out inside hot cells due to their high radioactivity

Post-Irradiation Examination and In-Pile Measurement Techniques for Water Reactor Fuels

International Nuclear Information System (INIS)

2009-12-01

in the 1960s when the construction of NPPs was being started. Evidently it can be assumed that infrastructure with basic unique equipments is old enough, both morally and physically, and needs to be up-graded or replaced. Thus, a sharp increase of the hydrocarbon fuel cost, green-house effect, necessity to construct more safe and efficient NPPs, justification of the lifetime prolongation of the existing NPPs, moral and physical ageing of the hot labs and research reactors equipment lead to the strong necessity to develop more perfect and more precise methods and equipment to examine irradiated components of nuclear reactors, first of all the most expensive one - nuclear fuel. Now the national hot laboratories and material testing reactors usually act as individual independent research establishments without any common and coordinated technical and business strategy towards the future needs and challenges. Even if there are not many joint programs for the development of nuclear power engineering in different countries, the method base and accumulated experience of the in- and post-reactor experiments should be widely shared so as to decrease the cost of this base in each country and to enforce its development. Thus, both problems and results of the application of new techniques to examine nuclear reactor components, as well as the conditions of separate labs should be discussed at the international level. The IAEA technical meetings are one of the most convenient means of arranging such discussion on the problems of the hot labs and research reactors development and application of new original techniques for examination of reactor materials properties. This publication represents a summary and proceedings of the two technical meetings (TMs) organized by IAEA on the subjects of Hot Cell Post-Irradiation Examination (PIE) Techniques and Pool Side Inspection of Water Reactor Fuel Assemblies and Fuel Rod Instrumentation and In-Pile Measurement Techniques. The first TM was

AGR-1 Fuel Compact 6-3-2 Post-Irradiation Examination Results

Energy Technology Data Exchange (ETDEWEB)

Paul demkowicz; jason Harp; Scott Ploger

2012-12-01

Destructive post-irradiation examination was performed on fuel Compact 6-3-2, which was irradiated in the AGR-1 experiment to a final compact average burnup of 11.3% FIMA and a time-average, volume-average temperature of 1070°C. The analysis of this compact was focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, measurement of fuel burnup by several methods, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy. A single particle with a defective SiC layer was identified during deconsolidation-leach-burn-leach analysis, which is in agreement with previous measurements showing elevated cesium in the Capsule 6 graphite fuel holder associated with this fuel compact. The fraction of the compact europium inventory released from the particles and retained in the matrix was relatively high (approximately 6E-3), indicating release from intact particle coatings. The Ag-110m inventory in individual particles exhibited a very broad distribution, with some particles retaining =80% of the predicted inventory and others retaining less than 25%. The average degree of Ag-110m retention in 60 gamma counted particles was approximately 50%. This elevated silver release is in agreement with analysis of silver on the Capsule 6 components, which indicated an average release of 38% of the Capsule 6 inventory from the fuel compacts. In spite of the relatively high degree of silver release from the particles, virtually none of the Ag-110m released was found in the compact matrix, and presumably migrated out of the compact and was deposited on the irradiation capsule components. Release of all other fission products from the particles appears to be less than a single

Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

International Nuclear Information System (INIS)

Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.

1981-04-01

Three fuel pin bundles, R-109/1, 2 and 3, were irradiated in a PWR loop in the HFR at Petten during respectively 131, 57 and 57 effective full power days at average powers of approximately 39 kW.m -1 and at peak powers of approximately 60 kW.m -1 . The results of the post-irradiation examinations of these fuel bundles are presented. (Auth.)

RERTR-12 Post-irradiation Examination Summary Report

Energy Technology Data Exchange (ETDEWEB)

Rice, Francine [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williams, Walter [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robinson, Adam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Meyer, Mitch [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabin, Barry [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-02-01

The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-property changes of the fuel and cladding.

Design, irradiation, and post-irradiation examination of the UC and (U,Pu)C fuel rods of the test groups Mol-11/K1 and Mol-11/K2

International Nuclear Information System (INIS)

Freund, D.; Elbel, H.; Steiner, H.

1976-06-01

The test groups K1 and K2 of the irradiation experiment Mol-11 are reported. Design, irradiation, and post-irradiation examination of the fuel rods irradiated are described. Mol-11/K1 consisted of one fuel rod with UC of 94% T.D. and helium bonding. This test group was intended to prove the high power irradiation capsule in pile. Mol-11/K2 consists of three fuel rods in total. One of these is presently still in the reactor. In this test group mixed carbide fuel of 83% T.D. and 15% Pu content under helium bonding is irradiated. The fuel rod K2-2 was provided with a capillary tube for the continuous measurement of fission gas pressure built up. 1.4988 stainless steel was chosen as cladding material. The final burnup lies between 35 and 70 MWd/kg M. Post-irradiation examination of the two test groups covers a theoretical analysis of the irradiation behaviour. (orig./GSCH) [de

Management of spent fuel from research and prototype power reactors and residues from post-irradiation examination of fuel

International Nuclear Information System (INIS)

1989-09-01

The safe and economic management of spent fuel is important for all countries which have nuclear research or power reactors. It involves all aspects of the handling, transportation, storage, conditioning and reprocessing or final disposal of the spent fuel. In the case of spent fuel management from power reactors the shortage of available reprocessing capacity and the rising economic interest in the direct disposal of spent fuel have led to an increasing interest in the long term storage and management of spent fuel. The IAEA has played a major role in coordinating the national activities of the Member States in this area. It was against this background that the Technical Committee Meeting on ''Safe Management of Spent Fuel From Research Reactors, Prototype Power Reactors and Fuel From Commercial Power Reactors That Has Been Subjected to PIE (Post Irradiated Examination)'' (28th November - 1st December 1988) was organised. The aims of the current meeting have been to: 1. Review the state-of-the-art in the field of management of spent fuel from research and prototype power reactors, as well as the residues from post irradiation examination of commercial power reactor fuel. The emphasis was to be on the safe handling, conditioning, transportation, storage and/or disposal of the spent fuel during operation and final decommissioning of the reactors. Information was sought on design details, including shielding, criticality and radionuclide release prevention, heat removal, automation and remote control, planning and staff training; licensing and operational practices during each of the phases of spent fuel management. 2. Identify areas where additional research and development are needed. 3. Recommend areas for future international cooperation in this field. Refs, figs and tabs

Facilities for post-irradiation examination of experimental fuel elements at Chalk River Nuclear Laboratories

International Nuclear Information System (INIS)

Mizzan, E.; Chenier, R.J.

1979-10-01

Expansion of post-irradiation facilities at the Chalk River Nuclear Laboratories and steady improvement in hot-cell techniques and equipment are providing more support to Canada's reactor fuel development program. The hot-cell facility primarily used for examination of experimental fuels averages a quarterly throughput of 40 elements and 110 metallographic specimens. New developments in ultrasonic testing, metallographic sample preparation, active storage, active waste filtration, and fissile accountability are coming into use to increase the efficiency and safety of hot-cell operations. (author)

Fuel fabrication and post-irradiation examination

Energy Technology Data Exchange (ETDEWEB)

Venter, P J; Aspeling, J C [Atomic Energy Corporation of South Africa Ltd., Pretoria (South Africa)

1990-06-01

This paper provides an overview of the A/c's Bevan and Eldopar facilities for the fabrication of nuclear fuel. It also describes the sophisticated Hot Cell Complex, which is capable of accommodating pressurised water reactor fuel and various other irradiated samples. Some interesting problems and their solutions are discussed. (author)

Fuel fabrication and post-irradiation examination

International Nuclear Information System (INIS)

Venter, P.J.; Aspeling, J.C.

1990-01-01

This paper provides an overview of the A/c's Bevan and Eldopar facilities for the fabrication of nuclear fuel. It also describes the sophisticated Hot Cell Complex, which is capable of accommodating pressurised water reactor fuel and various other irradiated samples. Some interesting problems and their solutions are discussed. (author)

Advanced post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2002-03-01

The purpose of the meeting was to provide and overview of the status of post-irradiation examination (PIE) techniques for water cooled reactor fuel assemblies and their components with emphasis given to advanced PIE techniques applied to high burnup fuel. Papers presented at the meeting described progress obtained in non-destructive (e.g. dimensional measurements, oxide layer thickness measurements, gamma scanning and tomography, neutron and X-ray radiography, etc.) and destructive PIE techniques (e.g. microstructural studies, elemental and isotopic analysis, measurement of physical and mechanical properties, etc.) used for investigation of water reactor fuel. Recent practice in high burnup fuel investigation revealed the importance of advanced PIE techniques, such as 3-D tomography, secondary ion mass spectrometry, laser flash, high resolution transmission and scanning electron microscopy, image analysis in microstructural studies, for understanding mechanisms of fuel behaviour under irradiation. Importance and needs for in-pile irradiation of samples and rodlets in instrumented rigs were also discussed. This TECDOC contains 20 individual papers presented at the meeting; each of the papers has been indexed separately

Available post-irradiation examination techniques at Romanian institute for nuclear research

International Nuclear Information System (INIS)

Parvan, Marcel; Sorescu, Antonius; Mincu, Marin; Uta, Octavian; Dobrin, Relu

2005-01-01

The Romanian Institute for Nuclear Research (INR) has a set of nuclear facilities consisting of TRIGA 14 MW(th) materials testing reactor and LEPI (Romanian acronym for post-irradiation examination laboratory) which enable to investigate the behaviour of the nuclear fuel and materials under various irradiation conditions. The available techniques of post-irradiation examination (PIE) and purposes of PIE for CANDU reactor fuel are as follows. 1) Visual inspection and photography by periscope: To examine the surface condition such as deposits, corrosion etc. 2) Eddy current testing: To verify the cladding integrity. 3) Profilometry and length measurement performed both before and after irradiation: To measure the parameters which highlight the dimensional changes i.e. diameter, length, diametral and axial sheath deformation, circumferential sheath ridging height, bow and ovality. 4) Gamma scanning and Tomography: To determine the burnup, axial and radial fission products activity distribution and to check for flux peaking and loading homogeneity. 5) Puncture test: To measure the pressure, volume and composition of fission gas and the inner free volume. 6) Optical microscopy: To highlight the structural changes and hydriding, to examine the condition of the fuel-sheath interface and to measure the oxide thickness and Vickers microhardness. 7) Mass spectrometry: To measure the burnup. 8) Tensile testing: To check the mechanical properties. So far, non-destructive and destructive post-irradiation examinations have been performed on a significant number of CANDU fuel rods (about 100) manufactured by INR and irradiated to different power histories in the INR 14 MW(th) TRIGA reactor. These examinations have been performed as part of the Romanian research programme for the manufacturing, development and safety of the CANDU fuel. The paper describes the PIE techniques and some results. (Author)

Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

Energy Technology Data Exchange (ETDEWEB)

Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D. [Chalk River Labs., Ontario (Canada)

1997-08-01

Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.

Hot Cell Post-Irradiation Examination and Poolside Inspection of Nuclear Fuel. Proceedings of the IAEA-HOTLAB Technical Meeting

International Nuclear Information System (INIS)

2013-04-01

The growing operational requirements for nuclear fuel, such as longer fuel cycles, higher burnups and wider use of transient regimes, require more robust fuel designs and more radiation resistant materials. Development of such advanced fuels is only possible with testing and analysis of their performance and application of adequate post-irradiation examination (PIE) methods and techniques. In addition, operational feedback data from poolside and PIE facilities are absolutely necessary for verification of fuel modelling codes and analysis of fuel failure mechanisms. For these reasons, the International Atomic Energy Agency (IAEA) has supported the international exchange of knowledge and sharing of best practices in the application of modern destructive and non-destructive methods of investigation of highly radioactive materials through a series of technical meetings (TMs), the last of which was held in 2006 in Buenos Aires. Since 1963, similar meetings, initially at the European level, have been organized by the Hot Laboratories and Remote Handling Working Group (HOTLAB), a partner in the development of the IAEA's Post Irradiation Examination Facilities Database (PIEDB), part of the IAEA's Integrated Nuclear Fuel Cycle Information System. With this successful partnership in mind, in 2010 the IAEA Technical Working Group on Fuel Performance and Technology recommended that a joint IAEA-HOTLAB TM be held on 'Hot Cell Post-Irradiation Examination and Pool-Side Inspection of Nuclear Fuel', covering questions relevant to the IAEA sub-programmes on 'Nuclear Power Reactor Fuel Engineering' and 'Management of Spent Fuel from Nuclear Power Reactors'. The TM was held on 23-27 May 2011, in Smolenice, Slovakia, with the participation of a large number of interested organizations and comprehensive coverage of major PIE and poolside inspection issues relating to both operation and storage of fuel for nuclear power reactors. The proceedings, summaries and conclusions of that joint

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-12-01

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-12-01

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).

Post-irradiation examination of U3SIX-AL fuel element manufactured and irradiated in Argentina

International Nuclear Information System (INIS)

Ruggirello, Gabriel; Calabroni, Hector; Sanchez, Miguel; Hofman, Gerard

2002-01-01

As a part of CNEA's qualification program as a supplier of low enriched Al-U 3 Si 2 dispersion fuel elements for research reactors, a post irradiation examination (PIE) of the first prototype of this kind, called P-04, manufactured and irradiated in Argentina, was carried out. The main purpose of this work was to set up various standard PIE techniques in the hot cell, looking forward to the next steps of the qualification program, as well as to acquire experience on the behaviour of this nuclear material and on the control of the manufacturing process. After an appropriate cooling period, on May 2000 the P-04 was transported to the hot cell in Ezeiza Atomic Centre. Non destructive and destructive tests were performed following the PIE procedures developed in Argonne National Laboratory (ANL), this mainly included dimensional measurement, microstructural observations and chemical burn-up analyses. The methodology and results of which are outlined in this report. The results obtained show a behaviour consistent with that of other fuel elements of the same kind, tested previously. On the other hand the results of this PIE, specially those concerning burn-up analysis and stability and corrosion behaviour of the fuel plates, will be of use for the IAEA Regional Program on the characterization of MTR spent fuel. (author)

Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

International Nuclear Information System (INIS)

Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

2009-01-01

The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

The second Euratom sponsored 9000C HTR fuel irradiation experiment in the HFR Petten Project E 96.02: Pt.2. Post-irradiation examination

International Nuclear Information System (INIS)

Roettger, R.; Bueger, J. de; Schoots, T.

1977-01-01

A large variety of HTR fuel specimens, loose coated particles, coupons and compacts provided by Belgonucleaire, the Dragon Project and the KFA Juelich have been irradiated in the HFR at Petten at about 900 0 C up to a maximum fast neutron fluence of about 7x10 21 cm -2 (EDN) as a Euratom sponsored experiment. The maximum burn-ups were between 11 and 18.5% FIMA. The results of the post-irradiation examinations, comprising visual inspection, dimensional measurements, microradiography, metallography, and burn-up determinations are presented in this part 2 of the final report. The examinations have shown that the endurance limit of most of the tested fuel varieties is beyond the reached irradiation values

Post-Irradiation Examination Test of the Parts of X-Gen Nuclear Fuel Assembly

International Nuclear Information System (INIS)

Ahn, S. B.; Ryu, W. S.; Choo, Y. S.

2008-08-01

The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this report are used to produce the irradiation data of the grid 1x1 cell spring, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 300 deg. C during about 100 days From the spring test of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. The tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor

Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

International Nuclear Information System (INIS)

Reitsamer, G.; Proksch, E.; Stolba, G.; Strigl, A.; Falta, G.; Zeger, J.

1985-01-01

Austrian R and D activities in the HTR-field reach back almost to the beginning of this advanced reactor line. For more than 20 years post-irradiation examination (PIE) of HTR-fuel has been performed at the laboratories of the Austrian Research Centre Seibersdorf Ltd. (OEFZS) (formerly OESGAE) and a high degree of qualification has been achieved in the course of that time. Most of the PIE-work has been carried out by international cooperation on contract basis with the OECD-DRAGON-project and with KFA-Juelich (FRG). There has also been some collaboration with GA (USA), Belgonucleaire and others in the past. HTR-fuel elements contain the fissile and fertile materials in form of coated particles (CPs) which are embedded in a graphite matrix. Because of this special design it has been necessary from the very beginning of the PIE work up to now to develop new methods (i.e. fuel element disintegration methods, chlorine gas leach, single particle examination techniques...) as well as to adapt and improve already existing methods (i.e. gamma spectrometry, mass-spectrometry, optical methods...). The main interests on PIE-work at Seibersdorf are concentrated on particle performance, fission product distribution and the 'free' Uranium content (contamination and broken particles) of the fuel elements (fuel spheres or cylindrical compacts). A short compilation of the applied methods and of available instrumental facilities is given as follows: deconsolidation of fuel elements; equipment for electrochemical deconsolidation; examinations and measurements of graphite and electrolyte samples; examination of coated particles; single particle examinations

Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

Energy Technology Data Exchange (ETDEWEB)

Reitsamer, G; Proksch, E; Stolba, G; Strigl, A; Falta, G; Zeger, J [Department of Chemistry, Austrian Research Centre Seibersdorf Ltd., Seibersdorf (Austria)

1985-07-01

Austrian R and D activities in the HTR-field reach back almost to the beginning of this advanced reactor line. For more than 20 years post-irradiation examination (PIE) of HTR-fuel has been performed at the laboratories of the Austrian Research Centre Seibersdorf Ltd. (OEFZS) (formerly OESGAE) and a high degree of qualification has been achieved in the course of that time. Most of the PIE-work has been carried out by international cooperation on contract basis with the OECD-DRAGON-project and with KFA-Juelich (FRG). There has also been some collaboration with GA (USA), Belgonucleaire and others in the past. HTR-fuel elements contain the fissile and fertile materials in form of coated particles (CPs) which are embedded in a graphite matrix. Because of this special design it has been necessary from the very beginning of the PIE work up to now to develop new methods (i.e., fuel element disintegration methods, chlorine gas leach, single particle examination techniques...) as well as to adapt and improve already existing methods (i.e. gamma spectrometry, mass-spectrometry, optical methods...). The main interests on PIE-work at Seibersdorf are concentrated on particle performance, fission product distribution and the 'free' Uranium content (contamination and broken particles) of the fuel elements (fuel spheres or cylindrical compacts). A short compilation of the applied methods and of available instrumental facilities is given as follows: deconsolidation of fuel elements; equipment for electrochemical deconsolidation; examinations and measurements of graphite and electrolyte samples; examination of coated particles; single particle examinations.

Fuel and fuel pin behaviour in a high burnup fast breeder fuel subassembly: Results of destructive post-irradiation examinations of the KNK II/1 fuel subassembly NY-205

International Nuclear Information System (INIS)

Patzer, G.

1991-05-01

The report gives a summarizing overview of the design characteristics, of the irradiation history and of the results of the destructive post-irradiation examinations of the fuel pins of the high-burnup fuel subassembly NY-205 of the KNK II first core. This element was operated for about 10 years and reached a maximum local burnup of 175 MWd/kg(HM) and a maximum neutron dose of 67 dpa-NRT. The main design data of this subassembly agree with those of the SNR 300 Mark-Ia, and it reached more than twice of the burnup and a similar neutron dose as foreseen for the SNR 300 fuel subassemblies [de

Integration of post-irradiation examination results of failed WWER fuel rods

International Nuclear Information System (INIS)

Smirnov, A.; Markov, D.; Smirnov, V.; Polenok, V.; Perepelkin, S.

2003-01-01

The aim of the work is to investigate the causes of WWER fuel rod failures and to reveal the dependence of the failed fuel rod behaviour and state on the damage characteristics and duration of their operation in the core. The post-irradiation examination of 12 leaky fuel assemblies (5 for WWER-440 and 7 for WWER-1000) has been done at SSC RF RIAR. The results show that the main mechanism responsible for the majority of cases of the WWER fuel rod perforation is debris-damage of the claddings. Debris fretting of the claddings spread randomly over the fuel assembly cross-section and they are registered in the area of the bundle supporting grid or under the lower spacer grids along the fuel assembly height. In the WWER fuel rods, the areas of secondary hydrogenating of cladding are spaced from the primary defects by ∼2500-3000 mm, as a rule, and are often adjacent closely to the upper welded joints. There is no pronounced dependence of the distance between the primary and secondary cladding defects neither on the linear power, at which the fuel rods were operated, nor on the period of their operation in the leaky state. The time period of the significant secondary damage formation is about 250 ± 50 calendar days for the WWER fuel rods with slight through primary defects (∼0.1 - 0.5 mm 2 ) operated in the linear power range 170-215 W/cm. Cladding degradation, taking place due to the secondary hydrogenating, does not occur in case of large through debris-defects during operation up to 600 calendar days

Operation of post-irradiation examination facility

Energy Technology Data Exchange (ETDEWEB)

Kim, E. G.; Jeon, Y. B.; Ku, D. S.

1996-12-01

In 1996, the post-irradiation examination(PIE) of nuclear fuels was performed as follows. It has been searched for the caution of defection of defected fuel rods of Youngkwang-4 reactor through NDT and metallographic examination that had been required by KEPCO. And in-pool inspection of Kori-1 spent fuel assembly(FO2) was carried out. HVAC system and pool water treatment system have been operated to maintain the facility safely, and electric power supply system was checked and maintained for the normal and steady supply electric power to the facility. Image processing software was developed for measurement of defection of spent fuel rods. Besides, a radiation shielding glove box was fabricated and a hot cell compressor for volume reduction of radioactive materials was fabricated and installed in hot cell. Safeguards of nuclear materials were implemented in strict accordance with the relevant Korean rules and regulations as well as the international non-proliferation regime. Also the IAEA inspection was carried out on the quarterly basis. (author). 31 tabs., 71 figs., 4 refs.

Post-irradiation examination of Oconee 1 fuel - cycle 1 destructive test phase

International Nuclear Information System (INIS)

1979-07-01

Standard B and W Mark-B (15 x 15) pressurized water reactor fuel rods were destructively examined after one cycle of irradiation in the Oconee 1 reactor. Fuel rod average burnup ranged from 10,603 to 11,270 MWd/mtU for the rods examined. Data obtained included fuel rod extraction loads, rod dimensional changes, cladding tensile properties, fuel pellet gap length, fission product distribution, fission gas and crud composition, fuel densification, chemical burnup analysis, and fuel and cladding microstructure. As expected, parametric changes were well within the design envelope. Superficial corrosion and wear were found at spacer grid contact points. However, the 19 rods examined were structurally sound and exhibited no indications of cladding defects associated with pelletcladding interactions

Advanced Post-Irradiation Examination Capabilities Alternatives Analysis Report

Energy Technology Data Exchange (ETDEWEB)

Jeff Bryan; Bill Landman; Porter Hill

2012-12-01

An alternatives analysis was performed for the Advanced Post-Irradiation Capabilities (APIEC) project in accordance with the U.S. Department of Energy (DOE) Order DOE O 413.3B, “Program and Project Management for the Acquisition of Capital Assets”. The Alternatives Analysis considered six major alternatives: ? No Action ? Modify Existing DOE Facilities – capabilities distributed among multiple locations ? Modify Existing DOE Facilities – capabilities consolidated at a few locations ? Construct New Facility ? Commercial Partnership ? International Partnerships Based on the alternatives analysis documented herein, it is recommended to DOE that the advanced post-irradiation examination capabilities be provided by a new facility constructed at the Materials and Fuels Complex at the Idaho National Laboratory.

Results of Microstructural Examinations of Irradiated LEU U-Mo Fuels

Energy Technology Data Exchange (ETDEWEB)

Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, ID 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organization (Australia)

2009-06-15

Introduction: The RERTR program is responsible for converting research reactors that use high-enriched uranium fuels to ones that use low-enriched uranium fuels [1]. As part of the development of LEU fuels, a variety of irradiation experiments are being conducted using the Advanced Test Reactor. Based on the results of initial fuel plate testing, adjustments have been made to the characteristics of fuel plates to improve the stability of the fuel microstructure. One improvement has been to add Si to the matrix of a dispersion fuel. This material is also being added at the fuel/cladding interface of a monolithic fuel. This paper will discuss the irradiation performance of these fuels, in terms of the stability of their microstructures during irradiation. Results and discussion: The post-irradiation examinations of fuel plates are performed at the Idaho National Laboratory. These examinations consist of visual examinations of fuel plates, gamma scanning, thickness measurements, oxide thickness measurements, and optical metallographic examinations of the fuel plate microstructures. Microstructural analysis is also performed using scanning electron microscopy. Overall, U-7Mo and U-10Mo alloy fuels have displayed the best irradiation performance, particularly, when a Si-containing Al alloy is used as the dispersion fuel matrix. The benefit of using this type of matrix is that the commonly observed fuel/cladding interaction that occurs during irradiation is reduced and the interaction layer that forms exhibit stable behavior during irradiation. Monolithic-type fuels, which consist of a U-Mo foil encased in Al alloy cladding, are also being developed. These types of fuels are also showing promise and will continue to be developed. One challenge with this type of fuel is in trying to maximize the bond strength at the foil/cladding interface. Fuel/cladding interactions can affect the quality of the boding at this interface. Si is being added to improve the characteristics

Post irradiation examination technology exchange

International Nuclear Information System (INIS)

Sozawa, Shizuo; Ito, Masayasu; Taguchi, Taketoshi; Nakagawa, Tetsuya; Lee, Hyung-Kwon

2012-01-01

Under the KAERI and JAEA agreement, in a part of the program 18 (Post Irradiation Examination (PIE) and Evaluation Technique of Irradiated Materials), an eddy current test was proposed as a round robin test, and it has been being progressed in both organizations in order to enhance the post irradiation examination technology. Up to now, several data are obtained by both PIE facilities. In this paper, the round robin test program is shown, and also shown obtained data with discussion from applicability as a nondestructive test in the hot cell. (author)

Post-irradiation examinations on the KNK II/1 fuel element NY-203 with 400 equivalent full-power days residence time and 10 % burnup

International Nuclear Information System (INIS)

Patzer, G.; Geier, F.

1984-09-01

The fuel assembly NY-203 has been irradiated in the first core of KNK II up to a burnup of about 10 % and a residence time of 400 equivalent full-power days. The assembly contained 211 fuel pins with 6.0 mm outer diameter and fuel pellets with the composition (U 0 .7Pu 0 .3)O 2 .00. The cladding material was the austenitic steel 1.4988 lg. Some selected pins were examined in the hot cells of the KfK Karlsruhe. The post-irradiation examinations did not reveal any critical design aspects [de

Technical review on irradiation tests and post-irradiation examinations in JMTR

International Nuclear Information System (INIS)

2017-07-01

The Japan Materials Testing Reactor (JMTR) has been contributing to various R and D activities in the nuclear research such as the fundamental research of nuclear materials/ fuels, safety research and development of power reactors, radio isotope (RI) production since its beginning of the operation in 1968. Irradiation technologies and post irradiation examination (PIE) technologies are the important factors for irradiation test research. Moreover, these technologies induce the breakthrough in area of nuclear research. JMTR has been providing unique capabilities for the irradiation test research for about 40 years since 1968. In future, any needs for irradiation test research used irradiation test reactors will continue, such as R and D of generation 4 power reactors, fundamental research of materials/fuels, RI production. Now, decontamination and new research reactor construction are common issue in the world according to aging. This situation is the same in Japan. This report outlines irradiation and PIE technologies developed at JMTR in 40 years to contribute to the technology transfer and human resource development. We hope that this report will be used for the new research rector design as well as the irradiation test research and also used for the human resource development of nuclear engineers in future. (author)

The operation of post-irradiation examination facility

International Nuclear Information System (INIS)

Kim, Eun Ka; Min, Duk Ki; Lee, Young Kil

1994-12-01

The operation of post-irradiation examination facility was performed as follow. HVAC and pool water treatment system were continuously operated, and radiation monitoring in PIE facility has been carried out to maintain the facility safely. Inspection of the fuel assembly (F02) transported from Kori Unit 1 was performed in pool, and fuel rods extracted from the fuel assembly (J44) of Kori Unit 2 NPP were examined in hot cell. A part of deteriorated pipe line of drinking water was exchanged for stainless steel pipe to prevent leaking accidents. Halon gas system was also installed in the exhausting blower room for fire fighting. And IAEA inspection camera for safeguard of nuclear materials was fixed at the wall in pool area. Radiation monitoring system were improved to display the area radioactive value at CRT monitor in health physics control room. And automatic check system for battery and emergency diesel generator was developed to measure the voltage and current of them. The performance test of oxide thickness measuring device installed in hot cell for irradiated fuel rod and improvement of the device were performed, and good measuring results using standard sample were obtained. The safeguard inspection of nuclear materials and operation inspection of the facility were carried out through the annual operation inspection, quarterly IAEA inspection and quality assurance auditing. 26 tabs., 43 figs., 14 refs. (Author) .new

International interest in the BONAPARTE measurement bench. Post-irradiation examination of lower-enriched fuel plates

International Nuclear Information System (INIS)

2014-01-01

The Belgian Nuclear Research Center SCK-CEN has developed a measurement bench (BONAPARTE) for the non-destructive analysis on fuel plate and rod type fuel elements. BONAPARTE is a modular device that can be employed for many purposes. The article discusses the employment of the BONAPARTE device for the accurate full post-irradiation mapping of fuel plate swelling with degree of precision of just a few micrometers.

AECL hot-cell facilities and post-irradiation examination services

International Nuclear Information System (INIS)

Schankula, M.H.; Plaice, E.L.; Woodworth, L.G.

1998-04-01

This paper presents an overview of the post-irradiation examination (PIE) services available at AECL's hot-cell facilities (HCF). The HCFs are used primarily to provide PIE support for operating CANDU power reactors in Canada and abroad, and for the examination of experimental fuel bundles and core components irradiated in research reactors at the Chalk River Laboratories (CRL) and off-shore. A variety of examinations and analyses are performed ranging from non-destructive visual and dimensional inspections to detailed optical and scanning electron microscopic examinations. Several hot cells are dedicated to mechanical property testing of structural materials and to determine the fitness-for-service of reactor core components. Facility upgrades and the development of innovative examination techniques continue to improve AECL's PIE capabilities. (author)

AECL hot-cell facilities and post-irradiation examination services

International Nuclear Information System (INIS)

Schankula, M.H.; Plaice, E.L.; Woodworth, L.G.

1995-01-01

This paper presents an overview of the post-irradiation examination (PIE) services available at AECL's hot-cell facilities (HCF). The HCFs are used primarily to provide PIE support for operating CANDU power reactors in Canada and abroad, and for the examination of experimental fuel bundles and core components irradiated in research reactors at the Chalk River Laboratories (CRL) and off-shore. A variety of examinations and analysis are performed ranging from non-destructive visual and dimensional inspections to detailed optical and scanning electron microscopic examinations. Several hot cells are dedicated to mechanical property testing of structural materials and to determine the fitness-for-service of reactor core components. Facility upgrades and the development of innovative examination techniques continue to improve AECL's PIE capabilities. (author)

Operation of post-irradiation examination facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Eun Ka; Park, Kwang Joon; Jeon, Yong Bum [and others; Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-02-01

In 1995, the post-irradiation examination (PIE) of nuclear fuels was performed as follows. The relation between burnup and top nozzle spring force of fuel assembly was obtained by measuring the holddown spring force on the Kori-1 reactor fuel assemblies. The resonance ultrasonic test for inspection of defect and moisture in fuel rod was carried out on fuel rods of C15 and J14 assemblies, and the change of fuel rod condition by storing in pool has been analyzed on the intentionally defected fuel rods (ID-C and ID-L) as well as intact fuel rod (1-2) by NDT in ht cell. The oxide layer thickness on cladding surface of J44-L12 fuel rod was measured by NDT method and metallography to reveal the oxidation as a function of temperature in the fuel rod, and the burnup of J44 fuel assembly was measured by chemical analysis. HVAC system and pool water treatment system of the PIE facility were continuously operated for air filtration and water purification. The monitoring of radiation and pool water in PIE facility has been carried out to maintain the facility safety, and electric power supply system was checked and maintained to supply the electric power to the facility normally. The developed measurement techniques of oxide layer thickness on fuel rod cladding and holddown spring force of top nozzle in fuel assembly were applied to examine the nuclear fuels. Besides, a radiation shielding glove box was designed and a hot cell compressor for volume reduction of radioactive materials was fabricated. 19 tabs., 38 figs., 7 refs. (Author) .new.

Post irradiation examinations of uranium-plutonium mixed carbide fuels irradiated at low linear power rate

International Nuclear Information System (INIS)

Maeda, Atsushi; Sasayama, Tatsuo; Iwai, Takashi; Aizawa, Sakuei; Ohwada, Isao; Aizawa, Masao; Ohmichi, Toshihiko; Handa, Muneo

1988-11-01

Two pins containing uranium-plutonium carbide fuels which are different in stoichiometry, i.e. (U,Pu)C 1.0 and (U,Pu)C 1.1 , were constructed into a capsule, ICF-37H, and were irradiated in JRR-2 up to 1.0 at % burnup at the linear heat rate of 420 W/cm. After being cooled for about one year, the irradiated capsule was transferred to the Reactor Fuel Examination Facility where the non-destructive examinations of the fuel pins in the β-γ cells and the destructive ones in two α-γ inert gas atmosphere cells were carried out. The release rates of fission gas were low enough, 0.44 % from (U,Pu)C 1.0 fuel pin and 0.09% from (U,Pu)C 1.1 fuel pin, which is reasonable because of the low central temperature of fuel pellets, about 1000 deg C and is estimated that the release is mainly governed by recoil and knock-out mechanisms. Volume swelling of the fuels was observed to be in the range of 1.3 ∼ 1.6 % for carbide fuels below 1000 deg C. Respective open porosities of (U,Pu)C 1.0 and (U,Pu)C 1.1 fuel were 1.3 % and 0.45 %, being in accordance with the release behavior of fission gas. Metallographic observation of the radial sections of pellets showed the increase of pore size and crystal grain size in the center and middle region of (U,Pu)C 1.0 pellets. The chemical interaction between fuel pellets and claddings in the carbide fuels is the penetration of carbon in the fuels to stainless steel tubes. The depth of corrosion layer in inner sides of cladding tubes ranged 10 ∼ 15 μm in the (U,Pu)C 1.0 fuel and 15 #approx #25 μm in the (U,Pu)C 1.1 fuel, which is correlative with the carbon potential of fuels posibly affecting the amount of carbon penetration. (author)

Post-irradiation examination of a fuel pin using a microscopic X-ray system: Measurement of carbon deposition and pin metrology

International Nuclear Information System (INIS)

Gras, Ch.; Stanley, S.J.

2008-01-01

The paper presents some interesting aspects associated with X-ray imaging and its potential application in the nuclear industry. The feasibility of using X-ray technology for the post-irradiation examination of a fuel pin has been explored, more specifically pin metrology and carbon deposition measurement. The non-active sample was specially designed to mimic the structure of an AGR fuel pin whilst a carbon based material was applied to the mock up fuel rod in order to mimic carbon deposition. Short duration low energy (50 kV) 2D digital radiography was employed and provided encouraging results (with respect to carbon deposition thickness and structure measurements) for the mock up fuel pin with a spatial resolution of around 10 μm. Obtaining quantitative data from the resultant images is the principal added value associated with X-ray imaging. A higher intensity X-ray beam (≥90 kV) was also used in conjunction with the low energy set-up to produce a clear picture of the cladding as well as the interface between the lead (Pb mimics the uranium oxide) and stainless steel cladding. Spent fuel metrology and routine radiography are two additional tasks that X-ray imaging could perform for the post-irradiation examination programme. Therefore, when compared to other techniques developed to deliver information on one particular parameter, X-ray imaging offers the possibility to extract useful information on a range of parameters

Status report on the irradiation testing and post-irradiation examination of low-enriched U3O8-Al and UAlsub(x)-Al fuel element by the Netherlands Energy Research Foundation (ECN)

International Nuclear Information System (INIS)

Pruimboom, H.; Lijbrink, E.; Otterdijk, K. von; Swanenburg de Veye, R.J.

1984-01-01

Within the framework of the RERTR-programme four low-enriched (20%) MTR-type fuel elements have been irradiated in the High Flux Reactor at Petten (The Netherlands) and are presently subjected to postirradiation examination. Two of the elements contain UAlsub(x)-Al and two contain U 3 O 8 -Al fuel. The test irradiation has been completed up to the target burn-up values of 50% and 75% respectively. An extensive surveillance programme carried out during the test period has confirmed the excellent in-reactor behaviour of both types. Post-irradiation examination of the 50% burn-up test elements, comprising of dimensional measurements, burn-up determination, fuel metallography and blister testing, has sofar confirmed the irradiation experiences. Good agreement between calculated and measured power and burn-up characteristics has been found. A survey of the test element characteristics, their irradiation history, the irradiation tests and the preliminary PIE results is given in the paper. (author)

The growth of intra-granular bubbles in post-irradiation annealed UO2 fuel

International Nuclear Information System (INIS)

White, R.J.

2001-01-01

Post-irradiation examinations of low temperature irradiated UO 2 reveal large numbers of very small intra-granular bubbles, typically of around 1 nm diameter. During high temperature reactor transients these bubbles act as sinks for fission gas atoms and vacancies and can give rise to large volumetric swellings, sometimes of the order of 10%. Under irradiation conditions, the nucleation and growth of these bubbles is determined by a balance between irradiation-induced nucleation, diffusional growth and an irradiation induced re-solution mechanism. This conceptual picture is, however, incomplete because in the absence of irradiation the model predicts that the bubble population present from the pre-irradiation would act as the dominant sink for fission gas atoms resulting in large intra-granular swellings and little or no fission gas release. In practice, large fission gas releases are observed from post-irradiation annealed fuel. A recent series of experiments addressed the issue of fission gas release and swelling in post-irradiation annealed UO 2 originating from Advanced Gas Cooled Reactor (AGR) fuel which had been ramp tested in the Halden Test reactor. Specimens of fuel were subjected to transient heating at ramp rates of 0.5 deg. C/s and 20 deg. C/s to target temperatures between 1600 deg. C and 1900 deg. C. The release of fission gas was monitored during the tests. Subsequently, the fuel was subjected to post-irradiation examination involving detailed Scanning Electron Microscopy (SEM) analysis. Bubble-size distributions were obtained from seventeen specimens, which entailed the measurement of nearly 26,000 intra-granular bubbles. The analysis reveals that the bubble densities remain approximately invariant during the anneals and the bubble-size distributions exhibit long exponential tails in which the largest bubbles are present in concentrations of 10 4 or 10 5 lower than the concentrations of the average sized bubbles. Detailed modelling of the bubble

The post-irradiation examination of fuel in support of Bruce A Nuclear Division fueling with flow program

International Nuclear Information System (INIS)

Montin, J.; Sagat, S.

1995-10-01

Bruce A Nuclear Division (BAND) units are operating at ∼ 75% of full power, because of the potential of a power pulse in the event of an inlet header break. As a result, BAND is converting to fueling with flow, to eliminate the potential of a power pulse and to allow for full-power operation. Concerns regarding the integrity of the end-of-life (EOL) bundles interacting with the latch at the downstream end of the fuel channel were raised. BAND carried out a test program in which EOL bundles in the upstream position of 13 of Unit 2 were cascaded into the downstream latch position 1 of another channel. Six of twelve cascaded bundles and two typical EOL position 13 (benchmark) bundles were selected for post-irradiation examination (PIE). Incipient cracks were found in the benchmark bundles. Metallographic and fractographic examination, along with crack dating, and hydrogen and deuterium analyses, indicated that the incipient cracks were the result of delayed-hydride assisted cracking at the EOL. Consequently, Ontario Hydro changed the design of the outlet shield plug to support all three rings of the fuel bundle, to minimize stress and prevent end plate cracking. Also, an ultrasonic end plate inspection tool (UT) was developed and located in the fuel bay, to inspect fuel-bundle end plates for cracks. A second test was done involving a series of four bundle cascades in BAND Unit 4 channels that had new outlet shield plugs. The latch bundles were discharged after a hot shutdown. The cascaded Unite 2 and Unit 4 latch bundles were checked for cracks using the UT. The PIE found incipient cracks or less-than-ideal welds in the assembly welds of fuel elements from Unit 2 (latch-supported fuel bundles) that had been identified by the UT as having incipient cracks. No incipient cracks were found in the assemble welds of fuel elements from Unit 4 (new outlet shield-supported fuel bundles) confirming the UT results. (author). 5 refs., 8 figs

Nondestructive post-irradiation examination of Loop-1, S1 and B1 rods

International Nuclear Information System (INIS)

Bratton, R.L.

1997-05-01

As a part of the Pacific Northwest National Laboratory's Tritium Target Development Program, eleven tritium target rods were irradiated in the Advanced Test Reactor located at the Idaho National Engineering and Environmental Laboratory during 1991. Both nondestructive and destructive post-irradiation examination on all eleven rods was planned under the Tritium Target Development Program. Funding for the program was reduced in 1991 resulting in the early removal of the program experiments before reaching their irradiation goals. Post-irradiation examination was only performed on one of the irradiated rods at the Pacific Northwest National Laboratory before the program was terminated in 1992. On December 6, 1995, the Secretary of Energy announced the pursuit of the Commercial Light-Water Reactor option for producing tritium establishing the Tritium Target Qualification Program at the Pacific Northwest National Laboratory. This program decided to pursue nondestructive and destructive post-irradiation examination of the ten remaining rods from the previous program. The ten rods comprise three experiments. The Loop-1 experiment irradiated eight target rods in a loop configuration for 217 irradiation days. The other two rods were irradiated in two separate irradiation experiments, designated as S1 and B1 for 143 effective full-power days, but at different power levels. After the ten rods were transferred from the ATR Canal to the Hot Fuels Examination Facility, the following examinations were performed: (1) visual examination and photography; (2) neutron radiography; (3) axial gamma scanning; (4) contact profilometry measurement; (5) bow and length measurements; (6) rod puncture and plenum gas analysis/measurement of plenum gas quantity; (7) void volume determination; and (8) internal pressure determination. This report presents the data collected during these examinations

The post irradiation examination of fuel in support of Bruce A nuclear division fueling with flow program

International Nuclear Information System (INIS)

Montin, J.; Sagat, S.; Day, R.; Novak, J.; Bromfield, H.

1995-01-01

Bruce A Nuclear Division (BAND) units are operating at ∼ 75% of full power, because of the potential of a power pulse in the event of an inlet header break. As a result, BAND is converting to fueling with flow, to eliminate the potential of a power pulse and to allow for full-power operation. Concerns regarding the integrity of the end-of-life (EOL) bundles interacting with the latch at the downstream end of the fuel channel were raised. BAND carried out a test program in which EOL bundles in the upstream position 13 of Unit 2 were cascaded into the downstream latch position 1 of another channel. Six of twelve cascaded bundles and two typical EOL position 13 (benchmark) bundles were selected for post-irradiation examination (PIE). Incipient cracks were found in the assembly welds (endplateto-endcap welds) of all six cascaded bundles. No incipient cracks were found in the benchmark bundles. Metallographic and fractographic examination, along with crack dating, and hydrogen and deuterium analyses, indicated that the incipient cracks were the result of delayed-hydride assisted cracking at the EOL. Consequently, Ontario Hydro changed the design of the outlet shield plug to support all three rings of the fuel bundle, to minimize stress and prevent endplate cracking. Also, an ultrasonic endplate inspection tool (UT) was developed and located in the fuel bay. to inspect fuelbundle endplates for cracks. A second test was done involving a series of four bundle cascades in BAND Unit 4 channels that had new outlet shield plugs. The latch bundles were discharged after a hot shutdown. The cascaded Unit 2 and Unit 4 latch bundles were checked for cracks using the UT. The PIE found incipient cracks or less-than-ideal welds in the assembly welds of fuel elements from Unit 2 (latch-supported fuel bundles) that had been identified by the UT as having incipient cracks. No incipient cracks were found in the assembly welds of fuel elements from Unit 4 (new outlet shield

IAEA Post Irradiation Examination Facilities Database

International Nuclear Information System (INIS)

Jenssen, Haakon; Blanc, J.Y.; Dobuisson, P.; Manzel, R.; Egorov, A.A.; Golovanov, V.; Souslov, D.

2005-01-01

The number of hot cells in the world in which post irradiation examination (PIE) can be performed has diminished during the last few decades. This creates problems for countries that have nuclear power plants and require PIE for surveillance, safety and fuel development. With this in mind, the IAEA initiated the issue of a catalogue within the framework of a coordinated research program (CRP), started in 1992 and completed in 1995, under the title of ''Examination and Documentation Methodology for Water Reactor Fuel (ED-WARF-II)''. Within this program, a group of technical consultants prepared a questionnaire to be completed by relevant laboratories. From these questionnaires a catalogue was assembled. The catalogue lists the laboratories and PIE possibilities worldwide in order to make it more convenient to arrange and perform contractual PIE within hot cells on water reactor fuels and core components, e.g. structural and absorber materials. This catalogue was published as working material in the Agency in 1996. During 2002 and 2003, the catalogue was converted to a database and updated through questionnaires to the laboratories in the Member States of the Agency. This activity was recommended by the IAEA Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT) at its plenary meeting in April 2001. The database consists of five main areas about PIE facilities: acceptance criteria for irradiated components; cell characteristics; PIE techniques; refabrication/instrumentation capabilities; and storage and conditioning capabilities. The content of the database represents the status of the listed laboratories as of 2003. With the database utilizing a uniform format for all laboratories and details of technique, it is hoped that the IAEA Member States will be able to use this catalogue to select laboratories most relevant to their particular needs. The database can also be used to compare the PIE capabilities worldwide with current and future

Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

Energy Technology Data Exchange (ETDEWEB)

Paul A Demkowicz; Paul Demkowicz; David V Laug

2010-10-01

Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

Program description for the qualification of CNEA - Argentina as a supplier of LEU silicide fuel and post-irradiation examinations plan for the first prototype irradiated in Argentina

International Nuclear Information System (INIS)

Rugirello, Gabriel; Adelfang, Pablo; Denis, Alicia; Zawerucha, Andres; Marco, Agustin di; Guillaume, Eduardo; Sbaffoni, Monica; Lacoste, Pablo

1998-01-01

In this report we present a description of the ongoing and future stages of the program for the qualification of CNEA, Argentina, as a supplier of low enriched uranium silicide fuel elements for research reactor. Particularly we will focus on the characteristics of the future irradiation experiment on a new detachable prototype, the post-irradiation examinations (PIE) plan for the already irradiated prototype PO4 and an overview of the recently implemented PIE facilities and equipment. The program is divided in several steps, some of which have been already completed. It concludes: development of the uranium silicide fissile material, irradiation and PIE of several full-scale prototypes. Important investments have been already carried out in the facilities for the FE production and PIE. (author)

Irradiation Behavior and Post-Irradiation Examinations of an Acoustic Sensor Using a Piezoelectric Transducer

International Nuclear Information System (INIS)

Lambert, T.; Zacharie-Aubrun, I.; Hanifi, K.; Valot, Ch.; Fayette, L.; Rosenkantz, E.; Ferrandis, J.Y.; Tiratay, X.

2013-06-01

The development of advanced instrumentation for in-pile experiments in Material Testing Reactor constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. In the framework of high burn-up fuel experiments under transient operating conditions, an innovative sensor based on acoustic method was developed by CEA and IES (Southern Electronic Institute).This sensor is used to determine the on-line composition of the gases located in fuel rodlet free volume and thus, allows calculating the molar fractions of fission gases and helium. The main principle of the composition determination by acoustic method consists in measuring the time of flight of an acoustic signal emitted and reflected in a specific cavity. A piezoelectric transducer, driven by a pulse generator, generates the acoustic wave in the cavity. The piezoelectric transducer is a PZT ceramic disk, mainly consisting of lead, zirconium and titanium. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. However, during the irradiation test, acoustic signal degradation was observed, mainly due to irradiation effect but also due to the increasing of the gas temperature. Despite this acoustic signal degradation, the time of flight measurements were carried out with good accuracy throughout the test, thanks to the development of a more efficient signal processing. After experiment, neutronic calculations were performed in order to determine neutron fluence at the level of the piezoelectric transducer. In order to have a better understanding of the acoustic sensor behavior under irradiation, Post Irradiation Examination program was done on piezoelectric transducer and on acoustic coupling material too. These examinations were also realized on a non-irradiated acoustic sensor built in the same conditions and with the same materials and the same

Post-irradiation handling and examination at the HFEF complex

International Nuclear Information System (INIS)

Bacca, J.P.

1980-01-01

The Hot Fuel Examination Facility provides postirradiation handling and examination of fast reactor irradiation experiments and safety tests for the United States Breeder Reactor Program. Nondestructive interim examinations and destructive terminal examinations at HFEF derive data from tests irradiated in the Experimental Breeder Reactor No. II, in the Transient Reactor Test Facility (TREAT), and in the Sodium Loop Safety Facility. Similar support will be provided in the near future for tests irradiated in the Fast Flux Test Facility, and for the larger sodium loops to be irradiated in TREAT

Post-irradiation examination of fifteen UO2/PuO2-fuel pins from the experiment DFR-350

International Nuclear Information System (INIS)

Geithoff, D.

1975-06-01

Within the framework of the fuel pin development for a sodium-cooled fast reactor a subassembly containing 77 fuel pins has been irradiated up to 5.65% fima in the Dounreay fast reactor. The pins were prototypes in terms of fuel and cladding material. The fuel consisted of mechanically mixed UO 2 (80%) and PuO 2 (20%) pressed into pellets whereas austenitic steels (W.-No. 1,4961 and 1,4988) were used as cladding material. Furthermore a blanket column of UO 2 pellets and a gas plenum were incorporated in the pin. For irradiation the conditions in a fast breeder were simulated by a linear rod power of 450 W/cm and a maximum cladding temperature of 630 0 C. After the successful completion of the irradiation, the subassembly was dismantled and fifteen pins were selected for a nondestructive and destructive examination. The tests included visual control, measurement of external dimensions, γ-spectroscopy, X-ray radiography, fission gas measurement, ceramography, radiochemical burn-up measurement. The results are presented. The most important results of the examinations seem to be the migration of fission product cesium and the fact that no signs of impending pin failure have been found. Thus the pin specification tested in this experiment is capable of achieving higher burnups under the irradiation conditions described above. (orig./AK) [de

Status of irradiation testing and PIE of MOX (Pu-containing) fuel

International Nuclear Information System (INIS)

Dimayuga, F.C.; Zhou, Y.N.; Ryz, M.A.

1995-01-01

This paper describes AECL's mixed oxide (MOX) fuel-irradiation and post-irradiation examination (PIE) program. Post-irradiation examination results of two major irradiation experiments involving several (U, Pu)O 2 fuel bundles are highlighted. One experiment involved bundles irradiated to burnups ranging fro 400 to 1200 MWh/kgHe in the Nuclear Power Demonstration (NPD) reactor. The other experiment consisted of several (U, Pu)O 2 bundles irradiated to burnups of up to 500 Mwh/kgHe in the National Research Universal (NRU) reactor. Results of these experiments demonstrate the excellent performance of CANDU MOX fuel. This paper also outlines the status of current MOX fuel irradiation tests, including the irradiation of various (U, Pu)O 2 bundles. The strategic importance of MOX fuel to CANDU fuel-cycle flexibility is discussed. (author)

Post irradiation examination and analysis of 13(U,Pu) C-fuel pins irradiated in the thermal flux of FR 2

International Nuclear Information System (INIS)

Weimar, P.; Steiner, H.

1979-01-01

The post-irradiation examination of the pins at Karlsruhe Hot Cells revealed the following results: Nearly all specimens showed noteworthy clad deformations (up to 3%). Defects in the form of cracks in the clad were found at three pins. The observed clad deformations resulted from mechanical interaction between fuel and cladding in consequence of an inexorable fuel swelling. A linear relationship between burnup and clad deformation was found. Defects were observed for burnups greater than 50 MWd/kgM and can be explained by the small fabrications clearances between clad and fuel pellets (50-90 μm) and high smear densities. Fission gas measurements were performed in a three fold way, gas release, gas trapped in pores and gas in solid solution in the lattice of the mixed carbide were determined. The gas release fraction showed values between 10 and 15%. Whereas the fission gas content trapped in large pores (> 1 μm) was linearly dependent on burnup, fission gas in small pores and in solid solution reached a saturation value at about 20 MWd/kgM. Measurements of micro-hardness revealed carburization depths of the clad of up to 40% at temperatures of about 650 0 C. Furtermore, it could be confirmed that the carburization depth followed an Arrhenius law. (orig.)

Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for Kijang research reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Man; Tahk, Young Wook; Jeong, Yong Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); and others

2017-08-15

The construction project of the Kijang research reactor (KJRR), which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm{sup 3}, was selected to achieve higher fuel efficiency and performance than are possible when using U{sub 3}Si{sub 2}/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1)], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm{sup 3}), were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination) results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for kijang research reactor

Directory of Open Access Journals (Sweden)

Jong Man Park

2017-08-01

Full Text Available The construction project of the Kijang research reactor (KJRR, which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm3, was selected to achieve higher fuel efficiency and performance than are possible when using U3Si2/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm3, were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

Nondestructive examination of irradiated fuel rods by pulsed eddy current techniques

International Nuclear Information System (INIS)

Francis, W.C.; Quapp, W.J.; Martin, M.R.; Gibson, G.W.

1976-02-01

A number of fuel rods and unfueled zircaloy cladding tubes which had been irradiated in the Saxton reactor have undergone extensive nondestructive and corroborative destructive examinations by Aerojet Nuclear Company as part of the Water Reactor Safety Research Program, Irradiation Effects Test Series. This report discusses the pulsed eddy current (PEC) nondestructive examinations on the fuel rods and tubing and the metallography results on two fuel rods and one irradiated zircaloy tube. The PEC equipment, designed jointly by Argonne National Laboratory and Aerojet, performed very satisfactorily the functions of diameter, profile, and wall thickness measurements and OD and ID surface defect detection. The destructive examination provided reasonably good confirmation of ''defects'' detected in the nondestructive examination

Development of cutting device for irradiated fuel rod

International Nuclear Information System (INIS)

Lee, E. P.; Jun, Y. B.; Hong, K. P.; Min, D. K.; Lee, H. K.; Su, H. S.; Kim, K. S.; Kwon, H. M.; Joo, Y. S.; Yoo, K. S.; Joo, J. S.; Kim, E. K.

2004-01-01

Post Irradiation Examination(PIE) on irradiated fuel rods is essential for the evaluation of integrity and irradiation performance of fuel rods of commercial reactor fuel. For PIE, fuel rods should be cut very precisely. The cutting positions selected from NDT data are very important for further destructive examination and analysis. A fuel rod cutting device was developed witch can cut fuel rods longitudinal very precisely and can also cut the fuels into the same length rod cuts repeatedly. It is also easy to remove the fuel cutting powder after cutting works and it can extend the life time of cutting device and lower the contamination level of hot cell

Behavior of pre-irradiated fuel under a simulated RIA condition

International Nuclear Information System (INIS)

Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide

1994-07-01

This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-3, conducted in the Nuclear Safety Research Reactor (NSSR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and analyses, interpretations, and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 19.6MWd/kgU with average linear heat rate of 25.3 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a deposited energy of 174±6 cal/g·fuel and a peak fuel enthalpy of 130±5 cal/g·fuel under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The cladding surface temperature increased to only 150degC, and the test resulted in slight fuel deformation and no fuel failure. An estimated rod-average fission gas release during the transient was about 2.2%. Through the detailed fuel examinations, the information concerning microstructural change in the fuel pellets were also obtained. (author)

The post irradiation examination of a sphere-pac (UPu)C fuel pin irradiated in the BR-2 reactor (MFBS 7 experiment)

International Nuclear Information System (INIS)

Smith, L.; Aerne, E.T.; Buergisser, B.; Flueckiger, U.; Hofer, R.; Petrik, F.

1979-09-01

A pin fuelled with Swiss made (UPu)C microspheres has been successfully irradiated to a peak burn-up of 6% fima in the Belgian BR2 Reactor. The pin, rated up to 95 kW/m, was intact after irradiation and exhibited a peak strain of just over 0.5%. The results of the post irradiation examination are reported. (Auth.)

HANARO fuel irradiation test(II)

Energy Technology Data Exchange (ETDEWEB)

Sohn, D. S.; Kim, H. R.; Chae, H. T.; Lee, B. C.; Lee, C. S.; Kim, B. G.; Lee, C. B.; Hwang, W

2001-04-01

In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiatied at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%.

Post-irradiation examination of the vipac fuel assemblies IFA-104 and IFA-203, irradiated in the Halden boiling water reactor

International Nuclear Information System (INIS)

Linde, A. van der; Lucas Luyckx, H.J.B.

1976-11-01

Two seven pin assemblies, IFA-104 and IFA-203, have been irradiated in the Halding Boiling Water Reactor. The Zircaloy clad pins contained vibrationally compacted, sharp edged UO 2 particles with smear densities of 84-88% theoretical density and a fuel stack length of about 1520 mm. The IFA-104 pins operated satisfactorily during approximately 240 effective full power days at an average linear heat rating of about 340 W/cm, and a peak rating of about 480 W/cm. Ingress of water via a pressure transducer into one of the IFA-203 pins necessitated the premature termination of the irradiation after approximately 30 effective full power days at an average rating of about 460 W/cm, and a peak rating of about 550 W/cm. One IFA-104 pin and two IFA-203 pins failed early. The primary cause of these failures has been sought in the oxidation reaction of the tantalum tube which protected the W/Re thermocouples with the surrounding UO 2 or with moisture in the fuel at temperatures in excess of approximately 1700degC. The fuel centre temperature and the internal gas pressure measurements during the beginning of lifetime and the results of the post-irradiation chemical and microscopic analysis on fuel pin cross-sections have been correlated with corresponding data calculated on a Gapcon-Thermal computer programme written for pellet fuel pins. Fairly good agreement could be achieved between the measured data on the vipac pins and the calculated data of a pellet pin. During the corrosion of the Zircaloy-2 cladding of the IFA-104 pins which were autoclaved in 400degC steam prior to the irradiation, relatively large amounts of H 2 entered the cladding from the D 2 O coolant side. With 0.66 mole% H 2 O in the D 2 O the calculated hydrogen uptake preference ratio was 33

Post-irradiation data analysis for NRC/PNL Halden assembly IFA-431

International Nuclear Information System (INIS)

Nealley, C.; Lanning, D.D.; Cunningham, M.E.; Hann, C.R.

1979-10-01

Results are presented for the post irradiation examination performed on IFA-431, which was a 6-rod test fuel assembly irradiated in Halden Reactor, Norway, under sponsorship of the Nuclear Regulatory Commission. The irradiation conditions included: peak powers of 33 kW/m; coolant pressure and temperature of 3.3 MPa and 240 0 C, respectively; and peak burnup of 4300 MWd/MTM. IFA-431 included instrumented rods of basic boiling water reactor design, with variations in fill gas composition, gap size, and UO 2 fuel type. The irradiation was designed to measure the effect of these variations upon fuel rod thermal and mechanical performance. The post irradiation examination assessed the permanent changes to the rods, including induced radioactivity, cladding deformation, fission gas release, and fuel densification

Post-Irradiation Examination of Fuel Pin R54-F20A, Irradiated in a NaK Environment. RCN Report

International Nuclear Information System (INIS)

Kwast, H.

1972-12-01

Fuel pin R54-F20A has been irradiated in a NaK-environment. Temperature measurements in the NaK were carried out at average linear fission powers of 552 and 825 W/cm respectively. A maximum average canning temperature of 920°C was reached. The fuel pin was irradiated for about 50 minutes at the maximum irradiation conditions, while the total irradiation time was two hours. The irradiation had to be broken off before the end condition was reached because of malfunctioning of the fuelfailure detection system. No power peaking did occur at the upper and lower interfaces between the 50%-enriched UO 2 - and the natural UO 2 + 8 w/o UB 4 pellet. About 35% of the fuel has molten, but the fuel pin did not fail. The irradiation has been carried out in the Poolside Facility (PSF) of the High Flux Reactor (HFR) at Petten. (author)

Examination of irradiated fuel elements using gamma scanning technique

International Nuclear Information System (INIS)

Ichim, O.; Mincu, M.; Man, I.; Stanica, M.

2016-01-01

The purpose of this paper is to validate the gamma scanning technique used to calculate the activity of gamma fission products from CANDU/TRIGA irradiated fuel elements. After a short presentation of the equipments used and their characteristics, the paper describes the calibration technique for the devices and how computed tomography reconstruction is done. Following the previously mentioned steps is possible to obtain the axial and radial profiles and the computed tomography reconstruction for calibration sources and for the irradiated fuel elements. The results are used to validate the gamma scanning techniques as a non-destructive examination method. The gamma scanning techniques will be used to: identify the fission products in the irradiated CANDU/TRIGA fuel elements, construct the axial and radial distributions of fission products, get the distribution in cross section through computed tomography reconstruction, and determine the nuclei number and the fission products activity of the irradiated CANDU/TRIGA fuel elements. (authors)

Examples of CEA managements of spent fuels from a prototype power reactor (PHENIX) and from commercial power reactors after post irradiation examinations

International Nuclear Information System (INIS)

Guay, P.

1988-01-01

CEA gained a good experience in the management of spent fuels from its research or power prototype reactors and of the fuel samples for post irradiation examinations. The solution for these products is the reprocessing. The delay to apply that solution is bound to the disponibility of the reprocessing facilities, and in several cases induce a delayed reprocessing. Only particular and limited fuels are planned to be sent in a definitive storage. The definitive storage is choosen only for a few fuels essentially requiring important modifications of the dissolution process. The treatments and operations on the spent fuels must be carried out following the French safety rules. Long and detailed flowsheet studies are therefore necessary before the setting up of the operations. Generally the cost of the management of limited quantities of fuels, as it is the case here, is high. The flowsheets are established in taking into account, as far as possible, the use of existing facilities, procedures, transport casks

HANARO fuel irradiation test (II): revision

Energy Technology Data Exchange (ETDEWEB)

Sohn, D. S.; Kim, H.; Chae, H. T.; Lee, C. S.; Kim, B. G.; Lee, C. B

2001-04-01

In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiated at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%. This report is the revision of KAERI/TR-1816/2001 on the irradiation test for HANARO fuel.

In-pile irradiation of rock-like oxide fuels

International Nuclear Information System (INIS)

Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H.

2001-01-01

Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO 2 single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

In-pile irradiation of rock-like oxide fuels

Energy Technology Data Exchange (ETDEWEB)

Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

2001-07-01

Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO{sub 2} single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

8 x 8 fuel surveillance program at Monticello site - end of Cycle 6: fourth post-irradiation inspection, October 1978

International Nuclear Information System (INIS)

Skarshaug, N.H.

1980-09-01

A fuel surveillance program for a lead 8 x 8 reload fuel assembly was implemented at the Monticello Nuclear Power Station in May 1974 prior to Reactor Cycle 3. Inspection results of the fourth post-irradiation inspection performed on this surveillance fuel assembly in October 1978 at EOC 6, after a bundle average exposure of 25,900 MWd/MT, are presented. The measurement techniques, results obtained and comparisons to previous measurements are discussed. The bundle and individual rods examined exhibited characteristics of normal operation and were approved for continued irradiation during Monticello operating Cycle 7

New facility for post irradiation examination of neutron irradiated beryllium

International Nuclear Information System (INIS)

Ishitsuka, Etsuo; Kawamura, Hiroshi

1995-01-01

Beryllium is expected as a neutron multiplier and plasma facing materials in the fusion reactor, and the neutron irradiation data on properties of beryllium up to 800 degrees C need for the engineering design. The acquisition of data on the tritium behavior, swelling, thermal and mechanical properties are first priority in ITER design. Facility for the post irradiation examination of neutron irradiated beryllium was constructed in the hot laboratory of Japan Materials Testing Reactor to get the engineering design data mentioned above. This facility consist of the four glove boxes, dry air supplier, tritium monitoring and removal system, storage box of neutron irradiated samples. Beryllium handling are restricted by the amount of tritium;7.4 GBq/day and 60 Co;7.4 MBq/day

Experimental and inspection facilities in post-irradiation of spent fuel pools for the analysis of the behaviour of nuclear fuels in power reactors

International Nuclear Information System (INIS)

Ruggirello, G.; Zawerucha, A.

1992-01-01

Since the beginning of the Atomic Nuclear Reactors (PHWR) Atucha I and Embalse in Argentine are employed different techniques for the knowing of the fuel bundles performances. It is detailed the facilities on post-irradiation examination. The techniques described are: online measurements, visual inspections, identifications of defective fuels and rods assemblies in spent fuel pools. This controls have made possible the feed-back to the manufactory process and the changes in the manufactory quality controls. (author)

Post-Irradiation Non-Destructive Analyses of the AFIP-7 Experiment

Science.gov (United States)

Williams, W. J.; Robinson, A. B.; Rabin, B. H.

2017-12-01

This article reports the results and interpretation of post-irradiation non-destructive examinations performed on four curved full-size fuel plates that comprise the AFIP-7 experiment. These fuel plates, having a U-10 wt.%Mo monolithic design, were irradiated under moderate operating conditions in the Advanced Test Reactor to assess fuel performance for geometries that are prototypic of research reactor fuel assemblies. Non-destructive examinations include visual examination, neutron radiography, profilometry, and precision gamma scanning. This article evaluates the qualitative and quantitative data taken for each plate, compares corresponding data sets, and presents the results of swelling analyses. These characterization results demonstrate that the fuel meets established irradiation performance requirements for mechanical integrity, geometric stability, and stable and predictable behavior.

Fabrication and post-irradiation examination of a zircaloy-2 clad UO2-1.5 wt% PuO2 fuel pin irradiated in PWL, CIRUS

International Nuclear Information System (INIS)

Sah, D.N.; Sahoo, K.C.; Chatterjee, S.; Majumdar, S.; Kamath, H.S.; Ramachandran, R.; Bahl, J.K.; Purushottam, D.S.C.; Ramakumar, M.S.; Sivaramakrishnan, K.S.; Roy, P.R.

1977-01-01

A zircaloy-2 clad UO 2 -1.5 wt% PuO 2 fuel pin was fabricated at the Radiometallurgy Section of the Bhabha Atomic Research Centre, Bombay, for irradiation in the pressurised water loop in CIRUS. Requisite development work related to powder conditioning, blending, pressing and sintering parameters was carried out to meet the exacting fuel pellet specifications of CANDU fuel. The fuel pin ruptured while being irradiated in the pressurised water loop in CIRUS, after experiencing a low burn-up of 507 MWD/MTM and was subsequently examined at the Radiometallurgy Hot Cells Facility. The results showed that internal clad hydriding led to primary failure of the fuel pin. Subsequent ingress of the coolant water caused excessive swelling of the thermal insulating magnesia pellets located at the ends of the fuel column. The swelling of magnesia pellets caused severe rupturing of the fuel pin at the two ends. The delayed rupturing of the fuel pin at the upper end, caused the fuel column to be displaced downwards by 5.85mm. (author)

Recent developments in post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting held in Cadarache, France, 17-21 October 1994

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

At the invitation of the Government of France, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) the IAEA convened a Technical Committee meeting from 14 to 21 October 1994 in Cadarache to discuss recent technical advances and improvements in the field of post-irradiation examination (PIE) of fuel used in nuclear power plants. Fifty participants representing 14 countries attended the meeting and 30 papers were presented and discussed during five technical sessions. Working Groups composed of the session chairmen and authors of papers prepared summaries of each session including conclusions and recommendations for future work. Refs, figs and tabs.

Recent developments in post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting held in Cadarache, France, 17-21 October 1994

International Nuclear Information System (INIS)

1995-09-01

At the invitation of the Government of France, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) the IAEA convened a Technical Committee meeting from 14 to 21 October 1994 in Cadarache to discuss recent technical advances and improvements in the field of post-irradiation examination (PIE) of fuel used in nuclear power plants. Fifty participants representing 14 countries attended the meeting and 30 papers were presented and discussed during five technical sessions. Working Groups composed of the session chairmen and authors of papers prepared summaries of each session including conclusions and recommendations for future work. Refs, figs and tabs

Post-irradiation analysis of low enriched U-Mo/Al dispersions fuel miniplate tests, RERTR 4 and 5

International Nuclear Information System (INIS)

Hofman, G.L.; Finlay, M.R.; Kim, Y.S.

2005-01-01

Interpretation of the post irradiation data of U-Mo/Al dispersion fuel mini plates irradiated in the Advanced Test Reactor to a maximum U-235 burn up of 80% are presented. The analyses addresses fuel swelling and porosity formation as these fuel performance issues relate to fuel fabrication and irradiation parameters. Specifically, mechanisms involved in the formation of porosity observed in the U-Mo/Al interaction phase are discussed and, means of mitigating or eliminating this irradiation phenomenon are offered. (author)

Irradiated fuel examination using the Cerenkov technique

International Nuclear Information System (INIS)

Nicholson, N.; Dowdy, E.J.

1981-03-01

A technique for monitoring irradiated nuclear fuel inventories located in water filled storage ponds has been developed and demonstrated. This technique provides sufficient qualitative information to be useful as a confirmatory technique to International Atomic Energy Agency inspectors. Measurements have been made on the Cerenkov glow light intensity from irradiated fuel that show the intensity of this light to be proportional to the cooling time. Fieldable instruments used in several tests confirm that such measurements can be made easily and rapidly, without fuel assembly movement or the introduction of apparatus into the storage ponds. The Cerenkov technique and instrumentation have been shown to be of potential use to operators of reactor spent fuel facilities and away from reactor storage facilities, and to the International Atomic Energy Agency inspectors who provide surveillance of the irradiated fuel stored in these facilities

Irradiation test and performance evaluation of DUPIC fuel

International Nuclear Information System (INIS)

Yang, Myung Seung; Song, K. C.; Moon, J. S.

2002-05-01

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

Experimental irradiation of UMo fuel: Pie results and modeling of fuel behaviour

International Nuclear Information System (INIS)

Languille, A.; Plancq, D.; Huet, F.; Guigon, B.; Lemoine, P.; Sacristan, P.; Hofman, G.; Snelgrove, J.; Rest, J.; Hayes, S.; Meyer, M.; Vacelet, H.; Leborgne, E.; Dassel, G.

2002-01-01

Seven full-sized U Mo plates containing ca. 8 g/cm 3 of uranium in the fuel meat have been irradiated since the beginning of the French U Mo development program. The first three of them with 20% 235 U enrichment were irradiated at maximum surfacic power under 150 W/cm 2 in the OSIRIS reactor up to 50% burn-up and are under examination. Their global behaviour is satisfactory: no failure and a low swelling. The other four plates were irradiated in the HFR Petten at maximum surfacic power between 150 and 250 W/cm 2 with two enrichments 20 and 35%. The experiment was stopped after two cycles due to a fuel failure. The post- irradiation examinations were completed in 2001 in Petten. Examinations showed a correct behaviour of 20% enriched plates and an abnormal behaviour of the two other plates (35%-enriched) with a clad failure on the plate 4. The fuel failure appears to result from a combination of factors that led to high corrosion cladding and high fuel meat temperatures. (author)

Development of a hot cell for post-irradiation analysis of nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Silva, Selma S.C.; Silva Junior, Silverio Ferreira da; Loureiro, Joao Roberto M. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: selmasallam@yahoo.com.br, e-mail: silvasf@cdtn.br, e-mail: jrmattos@cdtn.br

2009-07-01

Post irradiation examinations of nuclear fuels are performed in order to verify their in-service behavior. Examinations are conducted in compact structures called hot cells, designed to attend the different types of tests and analysis for fuel's characterization. The characterization of fuel microstructure is an activity performed in hot cells. Usually, hot cells for microstructural fuel analysis are designed to allow the metallographic and ceramographic samples preparation and after that, microscopical analysis of the fuel's microstructure. Due to the complexity of the foreseen operations, the severe limitations imposed by the available space into the hot cells, the capabilities of the remote manipulation devices, the procedures of radiological protection and the needs to obtain samples with an adequate surface quality for microscopic analysis, the design of a hot cell for fuel samples preparation presents a high level of complexity. In this paper, the methodology used to develop a hot cell facility for nuclear fuel's metallographic and ceramographic samples preparation is presented. Equipment, devices and systems used in conventional sample preparation processes were evaluated during bench tests. After the necessary adjustments and processes adaptations, they were assembled in a mock-up of the respective hot cell, where they were tested in conditions as realistic as possible, in order to improve the operations and processes to be performed at the real hot cells. (author)

Development of a hot cell for post-irradiation analysis of nuclear fuels

International Nuclear Information System (INIS)

Silva, Selma S.C.; Silva Junior, Silverio Ferreira da; Loureiro, Joao Roberto M.

2009-01-01

Post irradiation examinations of nuclear fuels are performed in order to verify their in-service behavior. Examinations are conducted in compact structures called hot cells, designed to attend the different types of tests and analysis for fuel's characterization. The characterization of fuel microstructure is an activity performed in hot cells. Usually, hot cells for microstructural fuel analysis are designed to allow the metallographic and ceramographic samples preparation and after that, microscopical analysis of the fuel's microstructure. Due to the complexity of the foreseen operations, the severe limitations imposed by the available space into the hot cells, the capabilities of the remote manipulation devices, the procedures of radiological protection and the needs to obtain samples with an adequate surface quality for microscopic analysis, the design of a hot cell for fuel samples preparation presents a high level of complexity. In this paper, the methodology used to develop a hot cell facility for nuclear fuel's metallographic and ceramographic samples preparation is presented. Equipment, devices and systems used in conventional sample preparation processes were evaluated during bench tests. After the necessary adjustments and processes adaptations, they were assembled in a mock-up of the respective hot cell, where they were tested in conditions as realistic as possible, in order to improve the operations and processes to be performed at the real hot cells. (author)

Metallographic examination of (uth) O2 and UO2 fuel tested in power ramp conditions in triga reactor

International Nuclear Information System (INIS)

Ioncescu, M.; Uta, O.

2015-01-01

The purpose of this paper is to determine the behavior of two fuel experimental elements (EC1 and EC2), by destructive post-irradiation examination. The fuel elements were mounted inside a pattern port, one in extension of the other and irradiated in power ramp conditions in order to check their behavior. Fuel element 1 (EC1) contains (UTh)O''2 pellet, and other one (EC2) UO''2 pellet. The results of destructive post-irradiation examination are evidenced by metallographic and ceramographic analyses. The data obtained from the post-irradiation examinations are used, first to confirm the security, reliability and nuclear fuel performance, and second, for the development of CANDU fuel. The results obtained by destructive examinations regarding the integrity, sheath hydrating and oxidation as well as the structural modifications are typical for fuel elements tested in power ramp conditions. (authors)

Postirradiation examination results for the Irradiation Effects Scoping Test 2

International Nuclear Information System (INIS)

Mehner, A.S.

1977-01-01

The postirradiation examination results are reported for two rods from the second scoping test (IE-ST-2) of the Nuclear Regulatory Commission Irradiation Effects Program. The rods were irradiated in the in-pile test loop of the Power Burst Facility at the Idaho National Engineering Laboratory. Rod IE-005 was fabricated from fresh fuel and cladding previously irradiated in the Saxton Reactor. Rod IE-006, fabricated from fresh fuel and unirradiated cladding, was equipped with six developmental cladding surface thermocouples. The rods were preconditioned, power ramped, and then subjected to film boiling operation. The performance of the rods and the developmental thermocouples are evaluated from the post irradiation examination results. The effects of prior irradiation damage in cladding are discussed in relation to fuel rod behavior during a power ramp and subsequent film boiling operation

Post-pulse detail metallographic examinations of low-enriched uranium silicide plate-type miniature fuel

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1991-10-01

Pulse irradiation at Nuclear Safety Research Reactor (NSRR) was performed using low-enriched (19.89 w% 235 U) unirradiated silicide plate-type miniature fuel which had a density of 4.8 gU/cm 3 . Experimental aims are to understand the dimensional stability and to clarify the failure threshold of the silicide plate-type miniature fuel under power transient conditions through post-pulse detail metallographic examinations. A silicide plate-type miniature fuel was loaded into an irradiation capsule and irradiated by a single pulse. Deposited energies given in the experiments were 62, 77, 116 and 154 cal/g·fuel, which lead to corresponding peak fuel plate temperatures, 201 ± 28degC, 187 ± 10degC, 418 ± 74degC and 871 ± 74degC, respectively. Below 400degC, reliability and dimensional stability of the silicide plate fuel was sustained, and the silicide plate fuel was intact. Up to 540degC, wall-through intergranular crackings occurred in the Al-3%Mg alloy cladding. With the increase of the temperature, the melting of the aluminum cladding followed by recrystallization, the denudation of fuel core and the plate-through intergranular cracking were observed. With the increase of the temperature beyond 400degC, the bowing of fuel plate became significant. Above the temperature of 640degC molten aluminum partially reacted with the fuel core, partially flowed downward under the influence of surface tension and gravity, and partially formed agglomerations. Judging from these experimental observations, the fuel-plate above 400degC tends to reduce its dimensional stability. Despite of the apparent silicide fuel-plate failure, neither generation of pressure pulse nor that of mechanical energy occurred at all. (J.P.N.)

Irradiation behavior of U 6Mn-Al dispersion fuel elements

Science.gov (United States)

Meyer, M. K.; Wiencek, T. C.; Hayes, S. L.; Hofman, G. L.

2000-02-01

Irradiation testing of U 6Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U 6Mn in an unrestrained plate configuration performs similarly to U 6Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3×10 27 m-3. Fuel plate failure occurs by fission gas pressure driven `pillowing' on continued irradiation.

Examination in hot laboratories of irradiated fuels from fast reactors

International Nuclear Information System (INIS)

Clottes, G.; Peray, R.; Ratier, J.L.

1980-05-01

Low irradiation rate examinations were carried out soon after the Rapsodie, Rapsodie Fortissimo and Phenix reactors were started up for the first time in order to check the level of maximum temperatures reached and the radial migration of oxygen and plutonium and to assess the movements of fuels inside the cladding. The other examinations were effected at a high specific burnup in order to defines the limit specific burnup securing the integrity of the fuel pin claddings (distortion, ruptures and possible consequences). The examinations carried out so far on fuel elements coming from Phenix or Rapsodie have allowed good fuel surveillance to be undertaken and the acquisition of a large number of data, thanks to which the fuel characteristics of future reactors of the system have been developed [fr

Irradiation Testing of TRISO-Coated Particle Fuel in Korea

International Nuclear Information System (INIS)

Kim, Bong Goo; Yeo, Sunghwan; Jeong, Kyung-Chai; Eom, Sung-Ho; Kim, Yeon-Ku; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung; Kim, Yong Wan

2014-01-01

In Korea, coated particle fuel is being developed to support development of a VHTR. At the end of March 2014, the first irradiation test in HANARO at KAERI to demonstrate and qualify TRISO-coated particle fuel for use in a VHTR was terminated. This experiment was conducted in an inert gas atmosphere without on-line temperature monitoring and control, or on-line fission product monitoring of the sweep gas. The irradiation device contained two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The duration of irradiation testing at HANARO was about 135 full power days from last August 2013. The maximum average power per particle was about 165 mW/particle. The calculated peak burnup of the TRISO-coated fuel was a little less than 4 atom percent. Post-irradiation examination is being carried out at KAERI’s Irradiated Material Examination Facility beginning in September of 2014. This paper describes characteristics of coated particle fuel, the design of the test rod and irradiation device for this coated particle fuel, and discusses the technical results of irradiation testing at HANARO. (author)

Irradiation of Argentine MOX fuels: Post-irradiation results and analysis

International Nuclear Information System (INIS)

Marino, A.C.; Perez, E.; Adelfang, P.

1997-01-01

The irradiation of the first Argentine prototypes of PHWR MOX fuels began in 1986. These experiments were made in the HFR-Petten reactor, Holland. The rods were prepared and controlled in the CNEA's facility. The postirradiation examinations were performed in the Kernforschungszentrum, Karlsruhe, Germany and in the JRC, Petten. The first rod has been used for destructive pre-irradiation analysis. The second one as a pathfinder to adjust systems in the HFR. Two additional rods including iodine doped pellets were intended to simulate 15000 MWd/T(M) burnup. The remaining two rods were irradiated until 15000 MWd/T(M) (BU15 experiment). One of them underwent a final ramp with the aim of verifying fabrication processes and studying the behaviour under power transients. BACO code was used to define the power histories and to analyze the experiments. This paper presents the postirradiation examinations for the BU15 experiments and a comparison with the BACO outputs for the rod that presented a failure during the ramp test of the BU15 experiment. (author). 17 refs, 30 figs, 5 tabs

Design and construction of the Fuels and Materials Examination Facility

International Nuclear Information System (INIS)

Burgess, C.A.

1979-01-01

Final design is more than 85 percent complete on the Fuels and Materials Examination Facility, the facility for post-irradiation examination of the fuels and materials tests irradiated in the FFTF and for fuel process development, experimental test pin fabrication and supporting storage, assay, and analytical chemistry functions. The overall facility is generally described with specific information given on some of the design features. Construction has been initiated and more than 10% of the construction contracts have been awarded on a fixed price basis

EDF requirements for hot cells examinations on irradiated fuel

International Nuclear Information System (INIS)

Segura, J.C.; Ducros, G.

2002-01-01

The objectives of increasing French Nuclear Power Plants (NPP) availability while lengthening the fuel irradiation cycle and reaching higher burnups lead EDF to carry out on site and hot cell examinations. The data issued from such fuel behaviour monitoring programmes will be used to ascertain that the design criteria are met. Data are also needed for modelling, development and validation. The paper deals quickly with the logistics linked to the selection and transport of fuel rods from NPP to hot cell laboratory. Hot cell PIEs remain a valuable method to obtain data in such fields as PCI (Pellet-Cladding Interaction), internal pressure, FGR (Fission Gas Release), oxide thickness, metallurgical aspects. The paper introduces burnup determination methods, inner pressure evaluation, preparation of samples for further irradiation such as power ramps for PCI and RIA (Reactivity Initiated Accident) testing. The nuclear microprobe of Perre Suee laboratory is also presented. (author)

Radiation protection aspects in the metallurgical examination of irradiated fuel elements

International Nuclear Information System (INIS)

Janardhanan, S.; Pillai, P.M.B.; Jacob, John; Kutty, K.N.; Wattamwar, S.B.; Mehta, S.K.

1981-01-01

The operational safety requirements of hot cell facilities for metallurgical examination of irradiated natural and enriched uranium fuel elements are highlighted. The cell shielding is designed for handling activities equivalent of 10 2 to 10 5 curies of gamma energy of 1.3 Mev. A brief outline of the built-in design features relevant to safety assessment is also incorporated. Reference is made to some salient features of Radiometallurgy Cells at Trombay. Metallurgical operations include investigations on cladding failure of irradiated material structure and specimen preparation from hot fuel element. The radiation protection aspects presented in this paper show that handling low irradiated fuel elements in these beta-gamma cells do not cause serious operational safety problems. The procedures followed and the containment provided would adequately restrict exposure of operational staff to acceptable limits. (author)

Behavior of pre-irradiated fuel under a simulated RIA condition. Results of NSRR Test JM-5

International Nuclear Information System (INIS)

Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Tanzawa, Sadamitsu; Ishijima, Kiyomi; Kobayashi, Shinsho; Kamata, Hiroshi; Homma, Kozo; Sakai, Haruyuki.

1995-11-01

This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-5, conducted in the Nuclear Safety Research Reactor (NSRR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and interpretations and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 25.7 MWd/kgU with average linear heat rate of 33.4 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a desposited energy of 223 ± 7 cal/g·fuel (0.93 ± 0.03 kJ/g·fuel) and a peak fuel enthalpy of 167 ± 5 cal/g·fuel (0.70 ± 0.02 kJ/g·fuel) under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The Test JM-5 resulted in cladding failure. More than twenty small cracks were found in the post-test cladding, and most of the defects located in pre-existing locally hydrided region. The result indicates an occurrence of fuel failure by PCMI (pellet/cladding mechanical interaction) in combination with decreased integrity of hydrided cladding. (author)

Over view of nuclear fuel cycle examination facility at KAERI

Energy Technology Data Exchange (ETDEWEB)

Lee, Key-Soon; Kim, Eun-Ga; Joe, Kih-Soo; Kim, Kil-Jeong; Kim, Ki-Hong; Min, Duk-Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-09-01

Nuclear fuel cycle examination facilities at the Korea Atomic Energy Research Institute (KAERI) consist of two post-irradiation examination facilities (IMEF and PIEF), one chemistry research facility (CRF), one radiowaste treatment facility (RWTF) and one radioactive waste form examination facility (RWEF). This paper presents the outline of the nuclear fuel cycle examination facilities in KAERI. (author)

Post-irradiation examination of the first SAP clad UO{sub 2} fuel elements irradiated in the X-7 organic loop

Energy Technology Data Exchange (ETDEWEB)

MacDonald, R. D.; Aspila, K.

1962-02-15

Seven fuel elements composing the first in-reactor test at Chalk River of SAP sheathing were irradiated in the X-7 organic loop. Activity, denoting a fuel failure, was detected in the loop coolant immediately after reactor start up; the fuel string was consequently removed from the loop nine hours later. Leak tests disclosed that five of the seven elements were defective. Inspection of the specimens showed essentially no change in element dimensions. Practically no organic fouling film was observed on the surface of the SAP cladding; organic coolant was found inside four of the defective elements. The appearance of the UO{sub 2} fuel was consistent with the irradiation time and the heat ratings achieved during the test. (author)

Post-irradiation examination of HTR-fuel at the Austrian Research Centre Seibersdorf Ltd

International Nuclear Information System (INIS)

Reitsamer, G.; Proksch, E.; Stolba, G.; Strigl, A.; Falta, G.; Zeger, J.

1984-02-01

This paper describes methods and measurements developed at the Austrian Research Centre Seibersdorf for the evaluation of the irradiation performance of HTR fuel. Main interest is concentrated on particle failure rates, fission product release, burn-up and inventory measurements (solid and gaseous fission products, uranium inventory). (Author) [de

Radiation protection aspects in the metallurgical examination of irradiated fuel elements

Energy Technology Data Exchange (ETDEWEB)

Janardhanan, S.; Pillai, P.M.B.; Jacob, J.; Kutty, K.N.; Wattamwar, S.B.; Mehta, S.K. (Bhabha Atomic Research Centre, Bombay (India). Health Physics Div.)

The operational safety requirements of hot cell facilities for metallurgical examination of irradiated natural and enriched uranium fuel elements are highlighted. The cell shielding is designed for handling activities equivalent of 10/sup 2/ to 10/sup 5/ curies of gamma energy of 1.3 Mev. A brief outline of the built-in design features relevant to safety assessment is also incorporated. Reference is made to some salient features of Radiometallurgy Cells at Trombay. Metallurgical operations include investigations on cladding failure of irradiated material structure and specimen preparation from hot fuel element. The radiation protection aspects presented in this paper show that handling low irradiated fuel elements in these beta-gamma cells do not cause serious operational safety problems. The procedures followed and the containment provided would adequately restrict exposure of operational staff to acceptable limits.

Fuel examination at SSEB Hunterston B power station

International Nuclear Information System (INIS)

Angell, I.; Oldfield, D.

1988-01-01

After a brief description of Hunterston 'B' Power Station and its fuel, the need for post irradiation examination is established. Means of providing this on site at various stages of the fuel route are described, i.e. refuelling machine, dismantling cell and storage pond. Techniques used include the human eye, video recording and endoscopy. (author)

Irradiation behavior of U{sub 6}Mn-Al dispersion fuel elements

Energy Technology Data Exchange (ETDEWEB)

Meyer, M.K. E-mail: mitchell.meyer@anl.gov; Wiencek, T.C.; Hayes, S.L.; Hofman, G.L

2000-04-01

Irradiation testing of U{sub 6}Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U{sub 6}Mn in an unrestrained plate configuration performs similarly to U{sub 6}Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3x10{sup 27} m{sup -3}. Fuel plate failure occurs by fission gas pressure driven 'pillowing' on continued irradiation.

Irradiation behaviors of coated fuel particles, (3)

International Nuclear Information System (INIS)

Fukuda, Kousaku; Kashimura, Satoru; Iwamoto, Kazumi; Ikawa, Katsuichi

1980-07-01

This report is concerning to the irradiation experiments of the coated fuel particles, which were performed by 72F-6A and 72F-7A capsules in JMTR. The coated particles referred to the preliminary design of VHTR were prepared for the experiments in 1972 and 1973. 72F-6A capsule was irradiated at G-10 hole of JMTR fuel zone for 2 reactor cycles, and 72F-7A capsule had been planned to be irradiated at the same irradiation hole before 72F-6A. However, due to slight leak of the gaseous fission products into the vacuum system controlling irradiation temperature, irradiation of 72F-7A capsule was ceased after 85 hrs since the beginning. In the post irradiation examination, inspection to surface appearance, ceramography, X-ray microradiography and acid leaching for the irradiated particle samples were made, and crushing strength of the two particle samples was measured. (author)

Post irradiation examinations cooperation and worldwide utilization of facilities

International Nuclear Information System (INIS)

Karlsson, Mikael

2009-01-01

Status of post irradiation examinations in Studsvik's facilities, cooperation and worldwide utilization of facilities, was described. Studsvik cooperate with irradiation facilities, as Halden, CEA and JAEA, as well as other hot cell facilities (examples, PSI, ITU and NFD) universities (example, the Royal Institute of Technology in Sweden) in order to be able to provide everything asked for by the nuclear community. Worldwide cooperation for effective use of expensive and highly specialized facilities is important, and the necessity of cooperation will be more and more recognized in the future. (author)

Post-irradiation examination of Oconee 1 fuel: end-of-cycle 2 nondestructive test phase

International Nuclear Information System (INIS)

1979-11-01

Standard B and W Mark B (15 x 15) pressurized water reactor fuel assemblies were nondestructively examined at the end of the second cycle of Oconee 1 reactor operation. Burnups of the 16 fuel assemblies examined ranged from 13,100 to 20,000 MWd/mtU. The examinations were conducted in the Oconee 1 and 2 spent fuel storage pool using the installed underwater test equipment. Data obtained included fuel rod and fuel assembly dimensions, water channel spacings, holddown spring forces, fuel rod crud characteristics, and fuel column axial gap and stack lengths. Visual examinations revealed no evidence of significant rod bowing, cladding deformation, cocked grids, or rod defects. The results, summarized in this report, indicate that the assemblies performed well through two cycles of reactor operation

The 3rd irradiation test plan of DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Yang, Myung Seung; Song, K. C.; Park, J. H. and others

2001-05-01

The objective of the 3rd irradiation test of DUPIC fuel at the HANARO is to estimate the in-core behaviour of a DUPIC pellet that is irradiated up to more than average burnup of CANDU fuel. The irradiation of DUPIC fuel is planned to start at May 21, 2001, and will be continued at least for 8 months. The burnup of DUPIC fuel through this irradiation test is thought to be more than 7,000 MWd/tHE. The DUPIC irradiation rig instrumented with three SPN detectors will be used to accumulate the experience for the instrumented irradiation and to estimate the burnup of irradiated DUPIC fuel more accurately. Under normal operating condition, the maximum linear power of DUPIC fuel was estimated as 55.06 kW/m, and the centerline temperature of a pellet was calculated as 2510 deg C. In order to assess the integrity of DUPIC fuel under the accident condition postulated at the HANARO, safety analyses on the locked rotor and reactivity insertion accidents were carried out. The maximum centerline temperature of DUPIC fuel was estimated 2590 deg C and 2094 deg C for each accident, respectively. From the results of the safety analysis, the integrity of DUPIC fuel during the HANARO irradiation test will be secured. The irradiated DUPIC fuel will be transported to the IMEF. The post-irradiation examinations are planned to be performed at the PIEF and IMEF.

Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation – Non-destructive analysis of the AFIP-1 fuel plates

Energy Technology Data Exchange (ETDEWEB)

Wachs, D.M., E-mail: daniel.wachs@inl.gov [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Robinson, A.B.; Rice, F.J. [Idaho National Laboratory, Characterization and Advanced PIE Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Kraft, N.C.; Taylor, S.C. [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Lillo, M. [Idaho National Laboratory, Nuclear Systems Design and Analysis Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Woolstenhulme, N.; Roth, G.A. [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2016-08-01

Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008–2009. The irradiation conditions were: ∼250 W/cm{sup 2} peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm{sup 3} peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

Release of fission products and post-pile creep behaviour of irradiated fuel rods stored under dry conditions

International Nuclear Information System (INIS)

Kaspar, G.; Peehs, M.; Bokelmann, R.; Jorde, D.; Schoenfeld, H.; Haas, W.; Bleier, A.; Rutsch, F.

1985-06-01

The release of moisture and fission products (Kr-85, H-3 and I-129) under dry storage conditions has been examined on six fuel rods which have become defective in the reactor. During the examinations, inert conditions prevailed and limited air inlet was allowed temporarily. The storage temperature was 400 0 C. The residual moisture content of the fuel rods was approx. 5 g. At the beginning of the test, the total moisture content and 0,05% (max.) of the fission gas inventory were released. Under inert conditions, fission gas was not released during a prolonged period of time. Under oxidizing conditions, however, fission gas was released in the course of UO 2 oxidation. Post-pile creep of Zircaloy cladding tubes was measured at temperatures between 350 and 395 0 C and interval gauge pressures between 69 and 110 bar. The creep curves indicate that the irradiated cladding tube specimens still bear internal residual stresses which contribute through their relaxation to the post-pile creep. (orig.) [de

Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment

Energy Technology Data Exchange (ETDEWEB)

Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts taken from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The fuel compacts examined in this study contained TRISO-coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.1% FIMA. These examinations are intended to explore kernel and coating morphology evolution during irradiation. This includes kernel porosity, swelling, and migration, and irradiation-induced coating fracture and separation. Variations in behavior within a specific cross section, which could be related to temperature or burnup gradients within the fuel compact, are also explored. The criteria for categorizing post-irradiation particle morphologies developed for AGR-1 ceramographic exams, was applied to the particles in the AGR-2 compacts particles examined. Results are compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.

Post Irradiation Capabilities at the Idaho National Laboratory

International Nuclear Information System (INIS)

Schulthess, J.L.; Rosenberg, K.E.

2011-01-01

The U.S. Department of Energy (DOE), Office of Nuclear Energy (NE) oversees the efforts to ensure nuclear energy remains a viable option for the United States. A significant portion of these efforts are related to post-irradiation examinations (PIE) of highly activated fuel and materials that are subject to the extreme environment inside a nuclear reactor. As the lead national laboratory, Idaho National Laboratory (INL) has a rich history, experience, workforce and capabilities for performing PIE. However, new advances in tools and techniques for performing PIE now enable understanding the performance of fuels and materials at the nano-scale and smaller level. Examination at this level is critical since this is the scale at which irradiation damage occurs. The INL is on course to adopt these advanced tools and techniques to develop a comprehensive nuclear fuels and materials characterization capability that is unique in the world. Because INL has extensive PIE capabilities currently in place, a strong foundation exist to build upon as new capabilities are implemented and work load increases. In the recent past, INL has adopted significant capability to perform advanced PIE characterization. Looking forward, INL is planning for the addition of two facilities that will be built to meet the stringent demands of advanced tools and techniques for highly activated fuels and materials characterization. Dubbed the Irradiated Materials Characterization Laboratory (IMCL) and Advanced Post Irradiation Examination Capability, these facilities are next generation PIE laboratories designed to perform the work of PIE that cannot be performed in current DOE facilities. In addition to physical capabilities, INL has recently added two significant contributors to the Advanced Test Reactor-National Scientific User Facility (ATR-NSUF), Oak Ridge National Laboratory and University of California, Berkeley.

Irradiation experiments of recycled PuO2-UO2 fuels by SAXTON reactor, (1)

International Nuclear Information System (INIS)

Yumoto, Ryozo; Akutsu, Hideo

1975-01-01

Seventy two mixed oxide fuel rods made by PNC were irradiated in Saxton Core 3. This paper generally describes the fuel specifications, the power history of the fuel and the post-irradiation examination of the PNC fuel. The specifications of the 4.0 w/o and 5.0 w/o enriched PuO 2 fuel rods with zircaloy-4 cladding are presented in a table and a figure. The positions of PNC fuel rods in the Saxton reactor are shown in a figure. Sixty eight 5.0 w/o PuO 2 -UO 2 fuel rods were assembled in a 9 x 9 rod array together with zircaloy-4 bars, a flux thimble, and a Sb-Be source. The power history of the Saxton Core 3 and the irradiation history of the PNC fuel rods are summarized in tables. The peak power and burnup of each fuel rod and the axial power profile are also presented. The maximum linear power rate and burnup attained were 512W/cm and 8700 MWD/T, respectively. As for the post irradiation examination, the items of nondestructive test, destructive test, and cladding test are presented together with the working flow diagram of the examination. It is concluded that the performance of all fuel rods was safe and satisfactory throughout the power history. (Aoki, K.)

Fabrication of MOX fuel element clusters for irradiation in PWL, CIRUS

International Nuclear Information System (INIS)

Roy, P.R.; Purushotham, D.S.C.; Majumdar, S.

1983-01-01

Three clusters, each containing 6 zircaloy-2 clad short length fuel elements of either MOX or UO 2 fuel pellets were fabricated for irradiation in pressurized water loop of CIRUS. The major objectives of the programme were: (a) to optimize the various fabrication parameters for developing a flow sheet for MOX fuel element fabrication; (b) to study the performance of the MOX fuel elements at a peak heat flux of 110 W/cm 2 ; and (c) to study the effect of various fuel pellet design changes on the behaviour of the fuel element under irradiation. Two clusters, one each of UO 2 and MOX, have been successfully irradiated to the required burn-up level and are now awaiting post irradiation examinations. The third MOX cluster is still undergoing irradiation. Fabrication of these fuel elements involved considerable amount of developing work related to the fabrication of the MOX fuel pellets and the element welding technique and is reported in detail in this report. (author)

Examination of the creep behaviour of ceramic fuel elements under neutron irradiation

International Nuclear Information System (INIS)

Brucklacher, D.

1978-01-01

This paper examines the creeping of UO 2 , UO 2 -PuO 2 and UN under neutron irradiation. It starts with the experimental results about the relation between the thermal creep rate and the load, the temperature, as well as characteristic material values, stoichiometry, grain size and porosity. These correlation are first qualitatively discussed and then compared with the statements of actual quantitative equations. From the models and theories on which these equations are based a modified Nabarro-Heering-equation results for the correlation between the creep rate of ceramic fuels, stress, temperature and the fission rate. In the experimental part of the examination, length-changes of creep samples of UO 2 , (U,Pu)O 2 and UN were measured in specially developed irradiation creep casings in different reactors. The measuring data were corrected and evaluated considering the thermal expansion effects, irregular temperature distribution and swelling effects in such a way that the dependences of the creep rate of UO 2 , UO 2 -PuO 2 and UN under irradiation on stress, temperature, fission rate, burn-up and porosity is obtained. It shows that creeping of fuels under irradiation at high temperatures is equivalent to thermally activated creeping, while at low temperature the creep rate induced by irradiation is much higher than the condition without irradiation. The increment of oxidic nuclear fuels is greater than in UN, the stress dependence on low burn-up is proportional in both cases, and the influence of temperature is quite small. (orig.) [de

Nondestructive examination of Oconee 1 fuel assemblies after four cycles of irradiation

International Nuclear Information System (INIS)

Pyecha, T.D.; Mayer, J.T.; Guthrie, B.A. III; Riordan, J.E.

1980-12-01

Five B and W Mark B (15 x 15) pressurized water reactor fuel assemblies were nondestructively examined after four cycles of irradiation in the Oconee 1 reactor. Four of the five assemblies examined had a burnup of 40,000 MWd/mtU; the fifth assembly had a burnup of 36,800 MWd/mtU. This effort is part of a Department of Energy program to improve uranium utilization by extending the burnup of light water reactor fuel. The examinations were conducted in the Oconee 1 and 2 spent fuel storage pool. Data obtained included fuel assembly and fuel rod dimensions, water channel spacings, spacer grid and holddown spring forces, fuel column stack and axial gap lengths, and crud samples. The results indicate that the assemblies performed well through four cycles of operation; all of the data were within design limits

Post irradiation examination of control rod assembly of FBTR

International Nuclear Information System (INIS)

Anandaraj, V.; Raghu, N.; Venkiteswaran, C.N.; Visweswaran, P.; Vijayakumar, Ran; Jayaraj, V.V.; Padmaprabu, P.; Saravanan, T.; Philip, John; Muralidharan, N.G.; Joseph, Jojo; Kasiviswanathan, K.V.

2010-01-01

Six control rods with boron carbide pellets are used in FBTR for shutdown and control of reactor power. One control rod after being subjected to a fluence level of 7.2 x 10 22 n/cm 2 was received for post irradiation examination (PIE) to assess its irradiation behavior and to investigate the incident of dropping of control rod. Examinations carried out include precise dimensional measurements to investigate the possibility of interference between the control rod and outer sheath, Neutron radiography and x-radiograph to assess the integrity of the boron carbide pellets and other internals, density measurements to assess the swelling behaviour of boron carbide pellets and metallographic examinations to study the cracking behaviour and microstructural changes in the pellet and the clad. Depletion of B 10 in the pellet was studied using time of flight mass spectrometry. The paper highlights the examinations and results of the PIE carried out. (author)

MOX fuel irradiation behavior in steady state (irradiation test in HBWR)

Energy Technology Data Exchange (ETDEWEB)

Kohno, S; Kamimura, K [Power Reactor and Nuclear Fuel Development Corp., Naka, Ibaraki (Japan)

1997-08-01

Two rigs of plutonium-uranium oxide (MOX) fuel rods have been irradiated in Halden boiling water reactor (HBWR) to investigate high burnup MOX fuel behavior for thermal reactor. The objective of irradiation tests is to investigate fuel behavior as influenced by pellet shape, pellet surface treatment, pellet-cladding gap size and MOX fuel powder preparations process. The two rigs have instrumentations for in-pile measurements of the fuel center-line temperature, plenum pressure, cladding elongation and fuel stack length change. The data, taken through in-operation instrumentation, have been analysed and compared with those from post-irradiation examination. The following observations are made: 1) PNC MOX fuels have achieved high burn-up as 59GWd/tMOX (67GWd/tM) at pellet peak without failure; 2) there was no significant difference in fission gas release fraction between PNC MOX fuels and UO{sub 2} fuels; 3) fission gas release from the co-converted fuel was lower than that from the mechanically blended fuel; 4) gap conductance was evaluated to decrease gradually with burn-up and to get stable in high burn-up region. 5) no evident difference of onset LHR for PCMI in experimental parameters (pellet shape and pellet-cladding gap size) was observed, but it decreased with burn-up. (author). 13 refs, 15 figs, 3 tabs.

Design of a decontamination section of the post-irradiation examination laboratory

International Nuclear Information System (INIS)

Homberger, Victor; Coronel, Ruben R.; Laumann, Victor; Perez, Jorge O.; Quinteros, Andrea N.; Ratner, Marcos

1999-01-01

The Post-Irradiation Examination Laboratory activities include the decontamination of expensive equipment of different sizes and weight, involving the complexity and extension of the necessary decontamination. A decontamination section has been designed for that purpose. (author)

Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

Energy Technology Data Exchange (ETDEWEB)

Paul Demkowicz; Scott Ploger; John Hunn

2012-05-01

The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Five irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These five compacts also included all four TRISO coating variations irradiated in the AGR experiment. The five compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. Approximately 40 to 80 particles within each cross section were exposed near enough to mid-plane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 830 classified particles allowed other relationships among morphological types to be established.

Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

International Nuclear Information System (INIS)

Demkowicz, Paul; Ploger, Scott; Hunn, John

2012-01-01

The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Five irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These five compacts also included all four TRISO coating variations irradiated in the AGR experiment. The five compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. Approximately 40 to 80 particles within each cross section were exposed near enough to mid-plane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 830 classified particles allowed other relationships among morphological types to be established.

Development status of post irradiation examination techniques at the JMTR Hot Laboratory

International Nuclear Information System (INIS)

Ohmi, M.; Ohsawa, K.; Nakagawa, T.; Umino, A.; Shimizu, M.; Satoh, H.; Oyamada, R.

1992-01-01

Hot laboratory at Oarai Research Establishment was founded to examine the objects mainly irradiated at JMTR (Japan Materials Testing Reactor) and has been operated since 1971. A wide variety of post-irradiation examinations (PIE) is available using the hot laboratory. Continuous efforts are made to develop new PIE techniques to accommodate the user's requirements. The following are main techniques recently developed in the hot laboratory; 1. Remote capsule assembly including remote weld of irradiated objects for reirradiation in JMTR. 2. Fracture toughness tests of reactor component materials. 3. Creep tests of heat resistance alloys in high temperature conditions. 4. Tests of irradiation assisted stress corrosion cracking (IASCC). 5. Examination techniques of miniaturized test specimens. This report describes an outline of the hot laboratory with main emphasis on the new PIE techniques. (author)

COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

OpenAIRE

ALEKSEY. L. IZHUTOV; VALERIY. V. IAKOVLEV; ANDREY. E. NOVOSELOV; VLADIMIR. A. STARKOV; ALEKSEY. A. SHELDYAKOV; VALERIY. YU. SHISHIN; VLADIMIR. M. KOSENKOV; ALEKSANDR. V. VATULIN; IRINA. V. DOBRIKOVA; VLADIMIR. B. SUPRUN; GENNADIY. V. KULAKOV

2013-01-01

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; th...

Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination

Science.gov (United States)

Collins, J. F.; Debogdan, C. E.; Diianni, D. C.

1973-01-01

A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.

Irradiation tests of THTR fuel elements in the DRAGON reactor (irradiation experiment DR-K3)

International Nuclear Information System (INIS)

Burck, W.; Duwe, R.; Groos, E.; Mueller, H.

1977-03-01

Within the scope of the program 'Development of Spherical Fuel Elements for HTR', similar fuel elements (f.e.) have been irradiated in the DRAGON reactor. The f.e. were fabricated by NUKEM and were to be tested under HTR conditions to scrutinize their employability in the THTR. The fuel was in the form of coated particles moulded into A3 matrix. The kernels of the particles were made of mixed oxide of uranium and thorium with an U 235 enrichment of 90%. One aim of the post irradiation examination was the investigation of irradiation induced changes of mechanical properties (dimensional stability and elastic behaviour) and of the corrosion behaviour which were compared with the properties determined with unirradiated f.e. The measurement of the fission gas release in annealing tests and ceramografic examinations exhibited no damage of the coated particles. The measured concentration distribution of fission metals led to conclusions about their release. All results showed, that neither the coated particles nor the integral fuel spheres experienced any significant changes that could impair their utilization in the THTR. (orig./UA) [de

Irradiation behaviors of coated fuel particles, (4)

International Nuclear Information System (INIS)

Fukuda, Kousaku; Kashimura, Satoru; Ogawa, Toru; Ikawa, Katsuichi; Iwamoto, Kazumi; Ishimoto, Kiyoshi

1981-09-01

Loose coated fuel particles prepared in confirmity to a preliminary design for the multi-purpose VHTR in fiscal 1972 - 1974 were irradiated by 73F - 12A capsule in JMTR. Main purpose for this irradiation experiment was to examine irradiation stability of the candidate TRISO coated fuel particles for the VHTR. Also the coated particles possessing low-density kernel (90%TD), highly anisotropic OLTI-PyC and ZrC coating layer were loaded with the candidate particles in this capsule. The coated particles were irradiated up to 1.5 x 10 21 n/cm 2 of fast neutron fluence (E > 0.18 MeV) and 3.2% FIMA of burnup. In the post irradiation examination it was observed that among three kinds of TRISO particles exposed to irradiation corresponding to the normal operating condition of the VHTR ones possessing poor characteristics of the coating layers did not show a good stability. The particles irradiated under abnormally high temperature condition (> 1800 0 C) revealed 6.7% of max. EOL failure fraction (95% confidence limit). Most of these particles were failed by the ameoba effect. Furthermore, among four kinds of the TRISO particles exposed to irradiation corresponding to the transient condition of the VHTR (--1500 0 C) the two showed a good stability, while the particles possessing highly anisotropic OLTI-PyC or poorly characteristic coating layers were not so good. (author)

Gamma spectrometrical examination of irradiated fuel

International Nuclear Information System (INIS)

Kristof, Edvard; Pregl, Gvido

1988-01-01

Gamma scanning is the only non-destructive technique for quantitative measuring of fission or activation products in spent fuel. The negligence of local variation of the linear attenuation coefficient of gamma rays in the irradiated fuel remains the main source of systematic error. To eliminate it we combine the (single) emission gamma ray scanning technique with a transmission measurement. Mathematical procedure joined with the experiment is particularly convenient for fuel elements of circular cross-section. In such a manner good results are obtainable even for relatively small number of measuring data. Accomplished routines enable to esteem the finite width of the collimation slit. The experiment has been partially automated. Trial measurements were carried out, and the measured data were successfully processed

The irradiation behavior of atomized U-Mo alloy fuels at high temperature

Science.gov (United States)

Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

2001-04-01

Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

INL Initial Input to the Mission Need for Advanced Post-Irradiation Examination Capability A Non-Major System Acquisition Project

International Nuclear Information System (INIS)

Tonc, Vince

2010-01-01

Consolidated and comprehensive post-irradiation examination (PIE) capabilities will enable the science and engineering understanding needed to develop the innovative nuclear fuels and materials that are critical to the success of the U.S. Department of Energy's (DOE) Office of Nuclear Energy (NE) programs. Existing PIE capabilities at DOE Laboratories, universities, and in the private sector are widely distributed, largely antiquated, and insufficient to support the long-range mission needs. In addition, DOE's aging nuclear infrastructure was not designed to accommodate modern, state-of-the-art equipment and instrumentation. Currently, the U.S. does not have the capability to make use of state-of-the-art technology in a remote, hot cell environment to characterize irradiated fuels and materials on the micro, nano, and atomic scale. This 'advanced PIE capability' to make use of state-of-the-art scientific instruments in a consolidated nuclear operating environment will enable comprehensive characterization and investigation that is essential for effectively implementing the nuclear fuels and materials development programs in support of achieving the U.S. DOE-NE Mission.

Status of fuel irradiation tests in HANARO

International Nuclear Information System (INIS)

Kim, Hark Rho; Lee, Choong Sung; Lee, Kye Hong; Jun, Byung Jin; Lee, Ji Bok

1999-01-01

Since 1996 after finishing the long-term operational test, HANARO (High-Flux Advanced Neutron Application Reactor) has been extensively used for material irradiation tests, beam application research, radioisotope production and neutron activation analysis. This paper presents the fuel irradiation test activities which are now conducted or have been finished in HANARO. KAERI developed LEU fuel using an atomization method for the research reactors. Using this LEU, we have set up and conducted three irradiation programs: (1) medium power irradiation test using a short-length mini-assembly made of 3.15 gU/cc U 3 Si, (2) high power irradiation tests using full-length test assemblies made of 3.15 gU/cc U 3 Si, and (3) irradiation test using a short-length mini-plate made of 4.8 gU/cc U 3 Si 2 . DUPIC (Direct Use of spent PWR fuels in CANDU Reactors) simulation fuel pellets, of which compositions are very similar to DUPIC pellets to keep the similarity in the thermo-mechanical property, were developed. Three mini-elements including 5 pellets each were installed in a capsule. This capsule has been irradiated for 2 months and unloaded from the HANARO core at the end of September 1999. Another very important test is the HANARO fuel qualification program at high power, which is required to resolve the licensing issue. This test is imposed on the HANARO operation license due to insufficient test data under high power environment. To resolve this licensing issue, we have been carrying out the required irradiation tests and PIE (Post-irradiation Examination) tests. Through this program, it is believed that the resolution of the licensing issue is achieved. In addition to these programs, several fuel test plans are under way. Through these vigorous activities of fuel irradiation test programs, HANARO is sure to significantly contribute to the national nuclear R and D programs. (author)

Some elaborating methods of gamma scanning results on irradiated nuclear fuels

International Nuclear Information System (INIS)

Sternini, E.

1979-01-01

Gamma scanning, as a post-irradiation examination, is a technique which provides a large number of informations on irradiated nuclear fuels. Power profile, fission products distribution, average and local burn-up of single elements structural and nuclear behaviour of fuel materials are examples of the obtained informations. In the present work experimental methods and theoretical calculations used at the CNEN hot cell laboratory for the mentioned purposes are described. Errors arising from the application of the gamma scanning technique are also discussed

Evaluation of fuel rods behavior - under irradiation test

International Nuclear Information System (INIS)

Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

1981-04-01

By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

Development of examination technique for oxide layer thickness measurement of irradiated fuel rods

International Nuclear Information System (INIS)

Koo, D. S.; Park, S. W.; Kim, J. H.; Seo, H. S.; Min, D. K.; Kim, E. K.; Chun, Y. B.; Bang, K. S.

1999-06-01

Technique for oxide layer thickness measurement of irradiated fuel rods was developed to measure oxide layer thickness and study characteristic of fuel rods. Oxide layer thickness of irradiated fuels were measured, analyzed. Outer oxide layer thickness of 3 cycle-irradiated fuel rods were 20 - 30 μm, inner oxide layer thickness 0 - 10 μm and inner oxide layer thickness on cracked cladding about 30 μm. Oxide layer thickness of 4 cycle-irradiated fuel rods were about 2 times as thick as those of 1 cycle-irradiated fuel rods. Oxide layer on lower region of irradiated fuel rods was thin and oxide layer from lower region to upper region indicated gradual increase in thickness. Oxide layer thickness from 2500 to 3000 mm showed maximum and oxide layer thickness from 3000 to top region of irradiated fuel rods showed decreasing trend. Inner oxide layer thicknesses of 4 cycle-irradiated fuel rod were about 8 μm at 750 - 3500 mm from the bottom end of fuel rod. Outer oxide layer thickness were about 8 μm at 750 - 1000 mm from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel. Oxide layer thickness technique will apply safety evaluation and study of reactor fuels. (author). 6 refs., 14 figs

HRB-22 capsule irradiation test for HTGR fuel. JAERI/USDOE collaborative irradiation test

Energy Technology Data Exchange (ETDEWEB)

Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

1998-03-01

As a JAERI/USDOE collaborative irradiation test for high-temperature gas-cooled reactor fuel, JAERI fuel compacts were irradiated in the HRB-22 irradiation capsule in the High Flux Isotope Reactor at the Oak Ridge National Laboratory (ORNL). Postirradiation examinations also were performed at ORNL. This report describes 1) the preirradiation characterization of the irradiation samples of annular-shaped fuel compacts containing the Triso-coated fuel particles, 2) the irradiation conditions and fission gas releases during the irradiation to measure the performance of the coated particle fuel, 3) the postirradiation examinations of the disassembled capsule involving visual inspection, metrology, ceramography and gamma-ray spectrometry of the samples, and 4) the accident condition tests on the irradiated fuels at 1600 to 1800degC to obtain information about fuel performance and fission product release behavior under accident conditions. (author)

Gamma scanning of mixed carbide and oxide fuel pins irradiated in FBTR

International Nuclear Information System (INIS)

Jayaraj, V.V.; Padalakshmi, M.; Ulaganathan, T.; Venkiteswaran, C.N.; Divakar, R.; Joseph, Jojo; Bhaduri, A.K.

2016-01-01

Fission in nuclear fuels results in a number of fission products that are gamma emitters in the energy range of 100 keV to 3 MeV. The gamma emitting fission products are therefore amenable for detection by gamma detectors. Assessment of the fission product distribution and their migration behavior through gamma scanning is important for characterizing the in reactor behavior of the fuel. Gamma scanning is an important non destructive technique used to evaluate the behavior of irradiated fuels. As a part of Post Irradiation Examinations (PIE), axial gamma scanning has been carried out on selected fuel pins of the FBTR Mark I mixed carbide fuel sub-assemblies and PFBR MOX test fuel sub-assembly irradiated in FBTR. This paper covers the results of gamma scanning and correlation of gamma scanning results with other PIE techniques

Microstructure and elemental distribution of americium containing MOX fuel under the short term irradiation tests

International Nuclear Information System (INIS)

Tanaka, Kosuke; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shin Ichi; Yoshimochi, Hiroshi; Tanaka, Kenya

2008-01-01

In order to investigate the effect of americium addition to MOX fuels on the irradiation behavior, the 'Am-1' program is being conducted in JAEA. The Am-1 program consists of two short term irradiation tests of 10-minute and 24 hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations (PIEs) are in progress. The PIEs for Am-containing MOX fuels focused on the microstructural evolution and redistribution behavior of Am at the initial stage of irradiation and the results to date are reported

An investigation on the irradiation behavior of atomized U-Mo/Al dispersion rod fuels

International Nuclear Information System (INIS)

Park, J.M.; Ryu, H.J.; Lee, Y.S.; Lee, D.B.; Oh, S.J.; Yoo, B.O.; Jung, Y.H.; Sohn, D.S.; Kim, C.K.

2005-01-01

The second irradiation fuel experiment, KOMO-2, for the qualification test of atomized U-Mo dispersion rod fuels with U-loadings of 4-4.5 gU/cc at KAERI was finished after an irradiation up to 70 at% U 235 peak burn-up and subjected to the IMEF (Irradiation material Examination Facility) for a post-irradiation analysis in order to understand the fuel irradiation performance of the U-Mo dispersion fuel. Current results for PIE of KOMO-2 revealed that the U-Mo/Al dispersion fuel rods exhibited a sound performance without any break-away swelling, but most of the fuel rods irradiated at a high linear power showed an extensive formation of the interaction phase between the U-Mo particle and the Al matrix. In this paper, the analysis of the PIE results, which focused on the diffusion related microstructures obtained from the optical and EPMA (Electron Probe Micro Analysis) observations, will be presented in detail. And a thermal modeling will be carried out to calculate the temperature of the fuel rod during an irradiation. (author)

Guidebook on destructive examination of water reactor fuel

International Nuclear Information System (INIS)

1997-01-01

As a result of common efforts of fuel vendors, utilities and research institutes the average burnup pf design batch fuels was increased for both PWRs and BWRs and the fuel failure rate has been reduced. The previously published Guidebook on Non-Destructive Examination of Water Reactor Fuel recommended that more detailed destructive techniques are required for complete understanding of fuel performance. On the basis of contributions of the 14 participants in the ED-WARF-II CRP and proceedings of IAEA Technical Committee on Recent Developments in Post-irradiation Examination Techniques for Water Reactor Fuel this guidebook was compiled. It gives a complete survey of destructive techniques available to date worldwide. The following examination techniques are described in detailed including major principles of equipment design: microstructural studies; elemental analysis; isotopic analysis; measurement of physical properties; measurement of mechanical properties. Besides the examination techniques, methods for refabrication of experimental rods from high burnup power reactor rods as well as methods for verification of non-destructive techniques by using destructive techniques is included

Irradiation behavior of experimental miniature uranium silicide fuel plates

International Nuclear Information System (INIS)

Hofman, Gerard L.; Neimark, L.A.; Mattas, R.F.

1983-01-01

Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk form, on the order of 7 x 10 20 cm -3 , far short of he approximately 20 x 10 20 cm -3 goal established for the RERTR Program. The purpose of the irradiation experiments on silicide fuels in the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix. The first group of experimental 'mini' fuel plates have recently reached the program's goal burnup and are in various stages of examination. Although the results to date indicate some limitations, it appears that within the range of parameters examined thus far the uranium silicide dispersion holds promise for satisfying most of the needs of the RERTR Program. The twelve experimental silicide dispersion fuel plates that were irradiated to approximately their goal exposure show the 30-vol % U 3 Si-Al plates to be in a stage of relatively rapid fission-gas-driven swelling at a fission density of 2 x 10 20 cm -3 . This fuel swelling will likely result in unacceptably large plate-thickness increases. The U 3 Si plates appear to be superior in this respect; however, they, too, are starting to move into the rapid fuel-swelling stage. Analysis of the currently available post irradiation data indicates that a 40-vol % dispersed fuel may offer an acceptable margin to the onset of unstable thickness changes at exposures of 2 x 10 21 fission/cm 3 . The interdiffusion between fuel and matrix

High burnup, high power irradiation behavior of helium-bonded mixed carbide fuel pins

International Nuclear Information System (INIS)

Levine, P.J.; Nayak, U.P.; Boltax, A.

1983-01-01

Large diameter (9.4 mm) helium-bonded mixed carbide fuel pins were successfully irradiated in EBR-II to high burnup (12%) at high power levels (100 kW/m) with peak cladding midwall temperatures of 550 0 C. The wire-wrapped pins were clad with 0.51-mm-thick, 20% cold-worked Type 316 stainless steel and contained hyperstoichiometric (Usub(0.8)Pusub(0.2))C fuel covering the smeared density range from 75-82% TD. Post-irradiation examinations revealed: extensive fuel-cladding mechanical interaction over the entire length of the fuel column, 35% fission gas release at 12% burnup, cladding carburization and fuel restructuring. (orig.)

Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

Energy Technology Data Exchange (ETDEWEB)

Venkiteswaran, C.N., E-mail: cnv@igcar.gov.in; Jayaraj, V.V.; Ojha, B.K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B.P.C.; Kasiviswanathan, K.V.; Jayakumar, T.

2014-06-01

The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel–clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel–clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

International Nuclear Information System (INIS)

Kurosaki, Ken; Uno, Masayoshi; Yamanaka, Shinsuke

1999-01-01

The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO 2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO 2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

Pre- and post-irradiation characterization and properties measurements of ZrC coated surrogate TRISO particles

Energy Technology Data Exchange (ETDEWEB)

Vasudevamurthy, Gokul [ORNL; Katoh, Yutai [ORNL; Hunn, John D [ORNL; Snead, Lance Lewis [ORNL

2010-09-01

Zirconium carbide is a candidate to either replace or supplement silicon carbide as a coating material in TRISO fuel particles for high temperature gas-cooled reactor fuels. Six sets of ZrC coated surrogate microsphere samples, fabricated by the Japan Atomic Energy Agency using the fluidized bed chemical vapor deposition method, were irradiated in the High Flux Isotope Reactor at the Oak Ridge National Laboratory. These developmental samples available for the irradiation experiment were in conditions of either as-fabricated coated particles or particles that had been heat-treated to simulate the fuel compacting process. Five sets of samples were composed of nominally stoichiometric compositions, with the sixth being richer in carbon (C/Zr = 1.4). The samples were irradiated at 800 and 1250 C with fast neutron fluences of 2 and 6 dpa. Post-irradiation, the samples were retrieved from the irradiation capsules followed by microstructural examination performed at the Oak Ridge National Laboratory's Low Activation Materials Development and Analysis Laboratory. This work was supported by the US Department of Energy Office of Nuclear Energy's Advanced Gas Reactor program as part of International Nuclear Energy Research Initiative collaboration with Japan. This report includes progress from that INERI collaboration, as well as results of some follow-up examination of the irradiated specimens. Post-irradiation examination items included microstructural characterization, and nanoindentation hardness/modulus measurements. The examinations revealed grain size enhancement and softening as the primary effects of both heat-treatment and irradiation in stoichiometric ZrC with a non-layered, homogeneous grain structure, raising serious concerns on the mechanical suitability of these particular developmental coatings as a replacement for SiC in TRISO fuel. Samples with either free carbon or carbon-rich layers dispersed in the ZrC coatings experienced negligible grain size

Nondestructive analysis of irradiated fuels

International Nuclear Information System (INIS)

Dudey, N.D.; Frick, D.C.

1977-01-01

The principal nondestructive examination techniques presently used to assess the physical integrity of reactor fuels and cladding materials include gamma-scanning, profilometry, eddy current, visual inspection, rod-to-rod spacing, and neutron radiography. LWR fuels are generally examined during annual refueling outages, and are conducted underwater in the spent fuel pool. FBR fuels are primarily examined in hot cells after fuel discharge. Although the NDE techniques are identical, LWR fuel examinations emphasize tests to demonstrate adherence to technical specification and reliable fuel performance; whereas, FBR fuel examinations emphasize aspects more related to the relative performance of different types of fuel and cladding materials subjected to variable irradiation conditions

POST-IRRADIATION ANALYSES OF U-MO DISPERSION FUEL RODS OF KOMO TESTS AT HANARO

Directory of Open Access Journals (Sweden)

H.J. RYU

2013-12-01

Full Text Available Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4–5 g-U/cm3 were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr, additional protective coatings (silicide or nitride, and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

Characterization of fission gas bubbles in irradiated U-10Mo fuel

Energy Technology Data Exchange (ETDEWEB)

Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.; Buck, Edgar C.

2017-09-01

Irradiated U-10Mo fuel samples were prepared with traditional mechanical potting and polishing methods with in a hot cell. They were then removed and imaged with an SEM located outside of a hot cell. The images were then processed with basic imaging techniques from 3 separate software packages. The results were compared and a baseline method for characterization of fission gas bubbles in the samples is proposed. It is hoped that through adoption of or comparison to this baseline method that sample characterization can be somewhat standardized across the field of post irradiated examination of metal fuels.

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

Energy Technology Data Exchange (ETDEWEB)

Boussard, F.; Huillery, R. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Combustibles; Averseng, J.L.; Serpantie, J.P. [Novatome Industries, 92 - Le Plessis-Robinson (France)

1994-12-31

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs.

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

International Nuclear Information System (INIS)

Boussard, F.; Huillery, R.

1994-01-01

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs

Equipment for detach the fuel elements of the irradiated candu fuel bundle

International Nuclear Information System (INIS)

Cojocaru, V.; Dinuta, G.

2013-01-01

Monitoring the behaviour of the fuel bundles during their combustion provides useful information for the operation of the nuclear power plant as well as for the fuel manufacturer. Before placing it inside the reactor, the fuel bundle is inspected visually, dimensionally and, during combustion in the reactor, its radioactive behaviour is monitored. The purpose of the presented equipment is to allow the visual external inspection of the damaged fuel bundle in order to identify visible defects and to detach the fuel element by breaking the welded connection between the cap and grid. These devices are operated using the handler devices already existing in the hot cells Post-Irradiation Examination Laboratory (LEPI). This equipment has been used successfully in the LEPI laboratory at SCN Pitesti to inspect the damaged fuel from Cernavoda NPP, in March 2013. (authors)

Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu

2008-01-01

U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during irradiation tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particle size was increases. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth. (author)

The KNK II/1 fuel assembly NY-205: Compilation of the irradiation history and the fuel and fuel pin fabrication data of the INTERATOM data bank system BESEX

International Nuclear Information System (INIS)

Patzer, G.; Geier, F.

1988-01-01

The fuel assembly NY-205 has been irradiated during the first and the second core of KNK II with a total residence time of 832 equivalent full-power days. A maximum burnup of 175.000 MWd/tHM or 18.6 % was reached with a maximum steel damage of 66 dpa-NRT. For the cladding the materials 1.4970 and 1.4981 have been used in different metallurgical conditions, and for the Uranium/Plutonium mixed- oxide fuel the most important variants of the major fabrication parameters had been realized. The assembly will be brought to the Hot Cells of the KfK Karlsruhe for post-irradiation examination in February 1988, so that the knowledge of the fabrication data is of interest for the selection of fuel pins and for the evaluation of the examination results. Therefore this report compiles the fuel and fuel pin fabrication data from the INTERATOM data bank system BESEX and additionally, an overview of the irradiation history of the assembly is given [de

AGR-1 Compact 4-1-1 Post-Irradiation Examination Results

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ploger, Scott A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); van Rooyen, Isabella J. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-02-01

Destructive post-irradiation examination was performed on AGR-1 fuel Compact 4-1-1, which was irradiated to a final compact-average burnup of 19.4% FIMA (fissions per initial metal atom) and a time-average, volume-average temperature of 1072°C. The analysis of this compact focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy, electron microscopy, and elemental analysis. Deconsolidation-leach-burn-leach (DLBL) analysis revealed no particles with failed TRISO or failed SiC layers (as indicated by very low uranium inventory in all of the leach solutions). The total fractions of the predicted compact inventories of fission products Ce-144, Cs-134, Cs-137, and Sr-90 that were present in the compact outside of the SiC layers were <2×10^-6, based on DLBL data. The Ag-110m fraction in the compact outside the SiC layers was 3.3×10^-2, indicating appreciable release of silver through the intact coatings and subsequent retention in the OPyC layers or matrix. The Eu-154 fraction was 2.4×10^-4, which is equivalent to the inventory in one average particle, and indicates a small but measurable level of release from the intact coatings. Gamma counting of 61 individual particles indicated no particles with anomalously low fission product retention. The average ratio of measured inventory to calculated inventory was close to a value of 1.0 for several fission product isotopes (Ce-144, Cs-134, and Cs-137), indicating good retention and reasonably good agreement with the predicted inventories. Measured-to-calculated (M/C) activity ratios for fission products Eu-154, Eu-155, Ru-106, Sb

Results of the post-irradiation examination of a highly-rated mixed oxide fuel rod from the Mol 7B experiment

International Nuclear Information System (INIS)

Coquerelle, M.; Walker, C.T.; Whitlow, W.H.

1980-01-01

The experiment MOL 7B was carried out in a epithermal flux in the Belgian reactor BR2. The pin examined contained fuel of initial composition (Usub(0.7)Pusub(0.3))Osub(1.98). It had been irradiated to a maximum burn-up of 13.2 at% at a maximum linear power of 568Wcm -1 . The fuel was clad with coldworked stainless steel. Electron microprobe analysis indicated that a Cr 2 O 3 type oxide was the main constituent of the grey phases in the gap. A metallic phase on the fuel surface had apparently resulted from the mechanical compaction of fragments of cladding that had been depleted in chromium by oxidation. Thus the main components of the phase were iron and nickel. Chromium loss from the inner cladding surface was significant only in the upper regions of the pin. In pin sections that were metallographically examined sigma phase and carbides of the type M 23 C 6 were present at the grain boundaries of the cladding. Cladding corrosion was not an Arrhenius function of the cladding temperature: the amount of metal lost from the inner cladding surface decreased with rise in cladding temperature above 910 K. A contributor to metal loss was the mechanical detachment of fragments of cladding which reformed as a metallic layer on the surface of the fuel. Chromium depletion and sigma phase formation at grain boundaries lowered the cohesive forces between grains which were then mechanically detached. Chromium loss from grain boundaries is mainly the results of oxidation of the cladding by the mixed oxide fuel. Data are presented to support the view that the local average O/M of the fuel determined the rate of oxidation and consequently the extent of chromium depletion. Fuel-cladding mechanical interactions were weak in the upper regions of the pin where metal loss was small

Development of U-Mo Research Reactor Fuel for Next Generation

International Nuclear Information System (INIS)

Park, Jong Man; Lee, Y. S.; Yang, J. H.; Ryu, H. J.; Kim, C. K.; Chae, H. T.; Seo, C. G.

2010-08-01

- Exportation of centrifugal atomized U-Mo powder - Completion of post irradiation examination for KOMO-3 irradiated fuel rods. - Select the dispersion fuel rod candidates for KOMO-4 irradiation test. - Irradiation test to solve the problems of interaction layer formation (KOMO-4) - Set the post irradiation examination of KOMO-4 irradiated fuel rods. - Development and characterization of innovative high U density fuel rods - Obtain and analyze foreign new irradiation test D

Description of the PIE facility for research reactors irradiated fuels in CNEA

International Nuclear Information System (INIS)

Bisca, A.; Coronel, R.; Homberger, V.; Quinteros, A.; Ratner, M.

2002-01-01

The PIE Facility (LAPEP), located at the Ezeiza Atomic Center (CAE), was designed to carry out destructive and non-destructive post-irradiation examinations (PIE) on research and power reactor spent fuels, reactor internals and other irradiated materials, and to perform studies related with: Station lifetime extension; Fuel performance; Development of new fuels; and Failures and determination of their causes. LAPEP is a relevant facility where research and development can be carried out. It is worth mentioning that in this facility the PIE corresponding to the Surveillance Program for the Atucha I Nuclear Power Plant (CNA-1) were successfully performed. Materials testing during the CNA-1 repair and the study of failures in fuel element plugs of the Embalse Nuclear Power Plant (CNE) were also performed. (author)

Extended burnup demonstration: reactor fuel program. Pre-irradiation characterization and summary of pre-program poolside examinations. Big Rock Point extended burnup fuel

International Nuclear Information System (INIS)

Exarhos, C.A.; Van Swam, L.F.; Wahlquist, F.P.

1981-12-01

This report is a resource document characterizing the 64 fuel rods being irradiated at the Big Rock Point reactor as part of the Extended Burnup Demonstration being sponsored jointly by the US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities. The program entails extending the exposure of standard BWR fuel to a discharge average of 38,000 MWD/MTU to demonstrate the feasibility of operating fuel of standard design to levels significantly above current limits. The fabrication characteristics of the Big Rock Point EBD fuel are presented along with measurement of rod length, rod diameter, pellet stack height, and fuel rod withdrawal force taken at poolside at burnups up to 26,200 MWD/MTU. A review of the fuel examination data indicates no performance characteristics which might restrict the continued irradiation of the fuel

Postirradiation examination of a low enriched U3Si2-Al fuel element manufactured and irradiated at Batan, Indonesia

International Nuclear Information System (INIS)

Suripto, A.; Sugondo, S.; Nasution, H.

1994-01-01

The first low-enriched U 3 Si 2 -Al dispersion plate-type fuel element produced at the Nuclear Fuel Element Center, BATAN, Indonesia, was irradiated to a peak 235 U burnup of 62%. Postirradiation examinations performed to data shows the irradiation behavior of this element to be similar to that of U 3 Si 2 -Al plate-type fuel produced and tested at other institutions. The main effect of irradiation on the fuel plates is a thickness increase of 30--40 μm (2.5-3.0%). This thickness increase is almost entirely due to the formation of a corrosion layer (Boehmite). The contribution of fuel swelling to the thickness increase is rather small (less than 10 μm) commensurate with the burnup of the fuel and the relatively moderate as-fabricated fuel volume fraction of 27% in the fuel meat

Post Irradiation Capabilities at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Schulthess, J.L.

2011-08-01

The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) oversees the research, development, and demonstration activities that ensure nuclear energy remains a viable energy option for the United States. Fuel and material development through fabrication, irradiation, and characterization play a significant role in accomplishing the research needed to support nuclear energy. All fuel and material development requires the understanding of irradiation effects on the fuel performance and relies on irradiation experiments ranging from tests aimed at targeted scientific questions to integral effects under representative and prototypic conditions. The DOE recently emphasized a solution-driven, goal-oriented, science-based approach to nuclear energy development. Nuclear power systems and materials were initially developed during the latter half of the 20th century and greatly facilitated by the United States ability and willingness to conduct large-scale experiments. Fifty-two research and test reactors with associated facilities for performing fabrication and pre and post irradiation examinations were constructed at what is now Idaho National Laboratory (INL), another 14 at Oak Ridge National Laboratory (ORNL), and a few more at other national laboratory sites. Building on the scientific advances of the last several decades, our understanding of fundamental nuclear science, improvements in computational platforms, and other tools now enable technological advancements with less reliance on large-scale experimentation.

Post Irradiation Capabilities at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Schulthess, J.L.; Robert D. Mariani; Rory Kennedy; Doug Toomer

2011-08-01

The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) oversees the research, development, and demonstration activities that ensure nuclear energy remains a viable energy option for the United States. Fuel and material development through fabrication, irradiation, and characterization play a significant role in accomplishing the research needed to support nuclear energy. All fuel and material development requires the understanding of irradiation effects on the fuel performance and relies on irradiation experiments ranging from tests aimed at targeted scientific questions to integral effects under representative and prototypic conditions. The DOE recently emphasized a solution-driven, goal-oriented, science-based approach to nuclear energy development. Nuclear power systems and materials were initially developed during the latter half of the 20th century and greatly facilitated by the United States’ ability and willingness to conduct large-scale experiments. Fifty-two research and test reactors with associated facilities for performing fabrication and pre and post irradiation examinations were constructed at what is now Idaho National Laboratory (INL), another 14 at Oak Ridge National Laboratory (ORNL), and a few more at other national laboratory sites. Building on the scientific advances of the last several decades, our understanding of fundamental nuclear science, improvements in computational platforms, and other tools now enable technological advancements with less reliance on large-scale experimentation.

Gamma-ray spectroscopy on irradiated fuel rods

International Nuclear Information System (INIS)

Terremoto, Luis Antonio Albiac

2009-01-01

The recording of gamma-ray spectra along an irradiated fuel rod allows the fission products to be qualitatively and quantitatively examined. Among all nondestructive examinations performed on irradiated fuel rods by gamma-ray spectroscopy, the most comprehensive one is the average burnup measurement, which is quantitative. Moreover, burnup measurements by means of gamma-ray spectroscopy are less time-consuming and waste-generating than burnup measurements by radiochemical, destructive methods. This work presents the theoretical foundations and experimental techniques necessary to measure, using nondestructive gamma-ray spectroscopy, the average burnup of irradiated fuel rods in a laboratory equipped with hot cells. (author)

Some Windscale experience of the underwater examination of water reactor fuel assemblies

International Nuclear Information System (INIS)

Banks, D.A.; Prestwood, J.; Stuttard, A.

1981-01-01

Windscale Nuclear Laboratories have been involved in the underwater post irradiation examination of irradiated water reactor fuel since the early 1970's. Since the work of the laboratories covers a wide range of fuel types, the equipment has had to be capable of handling any design, long or short, circular or square. There has so far been no element of routine work in the tasks performed at Windscale, for in this period fuel assemblies from 9 LWR's and WSGHWR have been examined successfully. Individual jobs have ranged from visual examination which may be carried out at several magnifications, to the complete breakdown of a PWR assembly to its separate rods and grids. Between these limits rod bow and rod diameter have been measured, rod withdrawal forces determined, and eddy current test methods devised. Cutting equipment has been used for a variety of dismantling tasks, and underwater cameras have been employed for monochrome and colour photography, using standard and macro lenses. The equipment is described. (author)

Irradiation performance of full-length metallic IFR fuels

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.

1992-07-01

An assembly irradiation of 169 full-length U-Pu-Zr metallic fuel pins was successfully completed in FFTF to a goal burnup of 10 at.%. All test fuel pins maintained their cladding integrity during the irradiation. Postirradiation examination showed minimal fuel/cladding mechanical interaction and excellent stability of the fuel column. Fission-gas release was normal and consistent with the existing data base from irradiation testing of shorter metallic fuel pins in EBR-II

The operation of post-irradiation examination facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Eun Ka; Park, Kwang Jun; Lee, Won Sang [and others; Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-01-01

The operation and management of PIE facility was executed in 1993. An indigenous 16 x 16 PWR type fuel assembly (ID No. J44) which was discharged from Kori unit 2 power reactor was transported to KAERI`s PIE facility and in-pool nondestructive examination and hot cell examination for the fuel were carried out. Because the above-mentioned 16 x 16 fuel is different from 14 x 14 fuel in its size and array of fuel rods, several examination and handling equipment for the 16 x 16 type fuel were designed and fabricated. PIE facility was operated in normal condition with the periodical check and inspection of the devices. The filter housing on the roof hood in chemical analysis hot cell was modified mounting air pressure gauge to indicate the optimal filter exchanging time. The burst air heating coil plate and the broken blowing fan of the HVAC system were repaired. The defaced grand packing in pool water circulation pump was replaced with the mechanical seal to prevent the leakage from the pump shaft sealing. The radiation monitoring in the facility was carried out to maintain the safe working condition and several radiation monitors were repaired. Spare parts for the radiation monitoring system were prepared to maintain the facility safely. The performance test of the emergency electric power supply system including UPS, battery and diesel generator was carried out. Oxide layer thickness measuring device for the performance test. Several devices including spent fuel handling equipment for the 17 x 17 PWR type fuel assembly were designed and fabricated for the subsequent PIE of nuclear fuels. 35 tabs., 17 figs., 7 refs. (Author) .new.

Study of candu fuel elements irradiated in a nuclear power plant

International Nuclear Information System (INIS)

Ionescu, S.; Uta, O.; Mincu, M.; Anghel, D.; Prisecaru, I.

2015-01-01

The object of this work is the behaviour of CANDU fuel elements after service in nuclear power plant. The results are analysed and compared with previous result obtained on unirradiated samples and with the results obtained on samples irradiated in the TRIGA reactor of INR Pitesti. Zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor, the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory (PIEL) of INR Pitesti on samples from fuel elements after they were removed out of the nuclear power plant: - dimensional and macrostructural characterization; - microstructural characterization by metallographic analyses; - determination of mechanical properties; - fracture surface analysis by scanning electron microscopy (SEM). A full set of non-destructive and destructive examinations concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the cladding was performed. The obtained results are typical for CANDU 6-type fuel. The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for the improvement of the CANDU fuel. (authors)

Gamma scanning of the irradiated HANARO fuels

International Nuclear Information System (INIS)

Hong, Kwon Pyo; Lee, K. S.; Park, D. G.; Baik, S. Y.; Song, W. S.; Kim, T. Y.; Seo, C. K.

1997-02-01

To conform the burnup state of the fuels, we have transported the irradiated HANARO fuels from the reactor to IMEF (Irradiated Material Examination Facility), and executed gamma scanning for the fuels. By measuring the gamma-rays from the irradiated fuels we could see the features of the relative burnup distributions in the fuel bundles. All of 17 fuel bundles were taken in and out between HANARO and IMEF from March till August in 1996, and we carried out the related regulations. Longitudinal gamma scanning and angular gamma scanning are done for each fuel bundle without dismantlement of the bundles. (author). 5 tabs., 25 figs

Microstructural evolution and Am migration behaviour in Am-containing fuels at the initial stage of irradiation

International Nuclear Information System (INIS)

Tanaka, Kosuke; Miwa, Shuhei; Sato, Isamu; Osaka, Masahiko; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shin-ichi; Yoshimochi, Hiroshi; Tanaka, Kenya

2010-01-01

In order to investigate the effect of americium addition to MOX fuels on the irradiation behaviour, the 'Am-1' programme is being conducted in JAEA. The Am-1 programme consists of two short-term irradiation tests of 10-minute and 24-hour irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post-irradiation examinations (PIE) are in progress. The PIE for Am-containing MOX fuels focused on the microstructural evolution and redistribution behaviour of Am at the initial stage of irradiation and the results to date are reported. The successful development of fabrication technology with remote handling and the evaluation of thermo-chemical properties based on the out-of-pile experiments are described with an emphasis on the effects of Am addition on the MOX fuel properties. (authors)

Pulsed eddy current inspection system for nondestructive examination of irradiated fuel rods

International Nuclear Information System (INIS)

Yancey, M.E.

1979-01-01

An inspection system has been developed for nondestructive examination of irradiated fuel rods utilizing pulsed eddy current techniques. The system employs an encircling type pulsed eddy current transducer capable of sensing small defects located on both the inner and outer diameter fuel rod surfaces during a single scan. Pulsed eddy current point probes are used to provide fuel rod wall thikness data and an indication of radial defect location. Two linear variable differential transformers are used to provide information on fuel rod diameter variation. A microprocessor based control system is used to automatically scan fuel rods up to 4.06 meters in length at predetermined radial locations. Defects as small as 0.005 cm deep by 0.254 cm long by 0.005 cm wide have been detected on outside diameter surfaces of a 1.43 cm outside diameter fuel rod cladding with a 0.094 cm wall thickness and 0.010 cm deep by 0.254 cm long by 0.005 cm wide on the inside diameter surface

Standard recommended practice for examination of fuel element cladding including the determination of the mechanical properties

International Nuclear Information System (INIS)

Anon.

1975-01-01

Guidelines are provided for the post-irradiation examination of fuel cladding and to achieve better correlation and interpretation of the data in the field of radiation effects. The recommended practice is applicable to metal cladding of all types of fuel elements. The tests cited are suitable for determining mechanical properties of the fuel elements cladding. Various ASTM standards and test methods are cited

Analysis of irradiation temperature in fuel rods of OGL-1 fuel assembly

International Nuclear Information System (INIS)

Fukuda, Kousaku; Kobayashi, Fumiaki; Minato, Kazuo; Ikawa, Katsuichi; Iwamoto, Kazumi

1984-10-01

Irradiation temperature in the fuel rods of 5th OGL-1 fuel assembly was analysed by the system composed by STPDSP2 and TRUMP codes. As the measured input-data, following parameters were allowed for; circumferential heating distribution around the fuel rod, which was measured in the JMTR critical assembly, axial heating distribution through the fuel rod, ratio of peak heatings of three fuel rods, and pre- and post-irradiation outer radii of the fuel compacts and inner radii of the graphite sleeves, which had been measured in PIE of the 5th OGL-1 fuel assembly. In computation the axial distributions of helium coolant temperature through the fuel rod and the heating value of each fuel rod were, firstly, calculated as input data for TRUMP. The TRUMP calculation yielded the temperatures which were fitted in those measured by all of the thermo-couples installed in the fuel rods, by adjusting only the value of the surface heat transfer coefficient, and consequently, the temperatures in all portions of the fuel rod were obtained. The apparent heat transfer coefficient changed to 60% of the initial values in the middle period of irradiation. For this reduction it was deduced that shoot had covered the surface of the fuel rod during irradiation, which was confirmed in PIE. Beside it, several things were found in this analysis. (author)

Post irradiation examination of type 316 stainless steels for in-pile Oarai water loop No.2 (OWL-2)

International Nuclear Information System (INIS)

Shibata, Akira; Kimura, Tadashi; Nagata, Hiroshi; Aoyama, Masashi; Kanno, Masaru; Ohmi, Masao

2010-11-01

The Oarai water loop No.2 (OWL-2) was installed in JMTR in 1972 for the purpose of irradiation experiments of fuel element and component material for light water reactors. Type 316 stainless steels (SSs) were used for tube material of OWL-2 in the reactor. But data of mechanical properties of highly irradiated Type 316 SSs has been insufficient since OWL-2 was installed. Therefore surveillance tests of type 316 SSs which were irradiated up to 3.4x10 25 n/m 2 in fast neutron fluence (>1 MeV) were performed. Meanwhile type 316 stainless steel (SS) is widely used in JMTR such as other irradiation apparatus and irradiation capsule, and additional data of type 316 SSs irradiated higher is required. Therefore post irradiation examinations of surveillance specimens made of type 316 SSs which were irradiated up to 1.0x10 26 n/m 2 in fast neutron fluence were performed and reported in this paper. In this result of surveillance tests of type 316 SSs irradiated up to 1.0x10 26 n/m 2 , tensile strength increase with increase of Neutron fluence and total elongation decreased with increase of Neutron fluence compared to unirradiated specimens and specimens irradiated up to 3.4x10 25 n/m 2 . This tendency has good agreement with results of 10 24 - 10 25 n/m 2 in fast neutron fluence. More than 37% in total elongation was confirmed in all test conditions. It was confirmed that type 316 SS irradiated up to 1.0x10 26 n/m 2 in fast neutron fluence has enough ductility as structure material. (author)

MICRO/NANO-STRUCTURAL EXAMINATION AND FISSION PRODUCT IDENTIFICATION IN NEUTRON IRRADIATED AGR-1 TRISO FUEL

Energy Technology Data Exchange (ETDEWEB)

van Rooyen, I. J.; Lillo, T. M.; Wen, H. M.; Hill, C. M.; Holesinger, T. G.; Wu, Y. Q.; Aguiara, J. A.

2016-11-01

Advanced microscopic and microanalysis techniques were developed and applied to study irradiation effects and fission product behavior in selected low-enriched uranium oxide/uranium carbide TRISO-coated particles from fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA. Although no TRISO coating failures were detected during the irradiation, the fraction of Ag-110m retained in individual particles often varied considerably within a single compact and at the capsule level. At the capsule level Ag-110m release fractions ranged from 1.2 to 38% and within a single compact, silver release from individual particles often spanned a range that extended from 100% retention to nearly 100% release. In this paper, selected irradiated particles from Baseline, Variant 1 and Variant 3 type fueled TRISO coated particles were examined using Scanning Electron Microscopy, Atom Probe Tomography; Electron Energy Loss Spectroscopy; Precession Electron Diffraction, Transmission Electron Microscopy, Scanning Transmission Electron Microscopy (STEM), High Resolution Electron Microscopy (HRTEM) examinations and Electron Probe Micro-Analyzer. Particle selection in this study allowed for comparison of the fission product distribution with Ag retention, fuel type and irradiation level. Nano sized Ag-containing features were predominantly identified in SiC grain boundaries and/or triple points in contrast with only two sitings of Ag inside a SiC grain in two different compacts (Baseline and Variant 3 fueled compacts). STEM and HRTEM analysis showed evidence of Ag and Pd co-existence in some cases and it was found that fission product precipitates can consist of multiple or single phases. STEM analysis also showed differences in precipitate compositions between Baseline and Variant 3 fuels. A higher density of fission product precipitate clusters were identified in the SiC layer in particles from the Variant 3 compact compared with the Variant 1 compact. Trend analysis shows

Test requirement for PIE of HANARO irradiated fuel rod

International Nuclear Information System (INIS)

Lim, I. C.; Cho, Y. G.

2000-06-01

Since the first criticality of HANARO reached in Feb. of 1995, the rod type U 3 Si-A1 fuel imported from AECL has been used. From the under-water fuel inspection which has been conducted since 1997, a ballooning-rupture type abnormality was observed in several fuel rods. In order to find the root cause of this abnormality and to find the resolution, the post irradiation examination(PIE) was proposed as the best way. In this document, the information from the under-water inspection as well as the PIE requirements are described. Based on the information in this document, a detail test plan will be developed by the project team who shall conduct the PIE

Post-Irradiation Examinations for Resolving Fuel Issues in Long Term Storage

International Nuclear Information System (INIS)

Karlsson, Joakim K.H.; Alvarez Holston, Anna-Maria

2014-01-01

In many countries extended long term dry storage is the solution for storage of spent nuclear fuel for the foreseeable future. The expected storage times have increased over the last years and today storage times of up to 300 years is anticipated. With such long storage times, requirements on transportability and retrievability of the fuel have become more important. Hitherto most investigations on fuel behaviour during dry storage have been focused on cladding creep and the impact of hydrogen and hydrides in the cladding. Creep data gives input to creep models and creep to rupture data helps to set criteria for maximum allowable internal rod pressure. Hydrides lower the ductility of the cladding and this is more pronounced with radially oriented hydrides. As the temperature decreases over time in a dry storage cask dissolved hydrogen will precipitate forming hydrides in addition to hydrides already present. Assuming there is sufficient hoop stress in the cladding, the new hydrides would be radially oriented. Together with lost ductility Delayed Hydride Cracking (DHC) could be a potential mechanism for rod failure over tens of years of dry storage as the temperature drops from about 350 deg. C to 150 deg. C. Hydride embrittlement and the DHC mechanism have been studied in the first Studsvik Cladding Integrity Project (SCIP), although the focus in this program has mainly been on higher temperatures relevant for operating conditions rather than on dry storage conditions. In addition to the mechanisms mentioned there are other failure mechanisms that could potentially threaten the cladding fuel integrity and retrievability. In case there is residual water or moisture available in the cask, or even in the fuel due to existing fuel failures, radiolysis gives free hydrogen and oxygen. In failed fuel this may cause fuel oxidation and swelling affecting fuel integrity. The hydrogen gas pressure will not threaten the cask but be available for cladding uptake. Furthermore

Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

Energy Technology Data Exchange (ETDEWEB)

Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

2013-12-15

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

International Nuclear Information System (INIS)

Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

2013-01-01

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60% 235 U; the mini-rods were irradiated to an average burnup of ∼ 85% 235 U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%

Irradiation test of fuel containing minor actinides in the experimental fast reactor Joyo

International Nuclear Information System (INIS)

Soga, Tomonori; Sekine, Takashi; Wootan, David; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi

2007-01-01

The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP, accounting for both prompt and delayed heating components, and then adjusted using E/C for 10 B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO 2-x or AmO 2-x in the (U, Pu)O 2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel. (author)

First results of the post-irradiation examination of the Ceramic Breeder materials from the Pebble Bed Assemblies Irradiation for the HCPB Blanket concept

International Nuclear Information System (INIS)

Hegeman, J.; Magielsen, A.J.; Peeters, M.; Stijkel, M.P.; Fokkens, J.H.; Laan, J.G. van der

2006-01-01

In the framework of developing the European Helium Cooled Pebble-Bed (HCPB) blanket an irradiation test of pebble-bed assemblies is performed in the HFR Petten. The experiment is focused on the thermo-mechanical behavior of the HCPB type breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. To achieve representative conditions a section of the HCPB is simulated by EUROFER-97 cylinders with a horizontal bed of ceramic breeder pebbles sandwiched between two beryllium beds. Floating Eurofer-97 steel plates separate the pebble-beds. The structural integrity of the ceramic breeder materials is an issue for the design of the Helium Cooled Pebble Bed concept. Therefore the objective of the post irradiation examination is to study deformation of pebbles and the pebble beds and to investigate the microstructure of the ceramic pebbles from the Pebble Bed Assemblies. This paper concentrates on the Post Irradiation Examination (PIE) of the four ceramic pebble beds that have been irradiated in the Pebble Bed Assembly experiment for the HCPB blanket concept. Two assemblies with Li 4 SiO 4 pebble-beds are operated at different maximum temperatures of approximately 600 o C and 800 o C. Post irradiation computational analysis has shown that both have different creep deformation. Two other assemblies have been loaded with a ceramic breeder bed of two types of Li 2 TiO 3 beds having different sintering temperatures and consequently different creep behavior. The irradiation maximum temperature of the Li 2 TiO 3 was 800 o C. To support the first PIE result, the post irradiation thermal analysis will be discussed because thermal gradients have influence on the pebble-bed thermo-mechanical behavior and as a result it may have impact on the structural integrity of the ceramic breeder materials. (author)

Bilateral cooperation between Germany and Brazil on fuel irradiation

International Nuclear Information System (INIS)

Dias, J.W.

1977-01-01

Within the framework of the Government Agreement on Scientific and Technical Cooperation between the Federal Republic of Germany and Brazil, the Brazilian National Atomic Commission and the Juelich Nuclear Research Center (KFA) signed on 23rd April, 1971 an Agreement on Cooperation in the field of Nuclear Research and Reactor Technology. Projects have been elaborated in fields of mutual interest to share activities between the partner institutes in both countries. A typical project is the fuel irradiation programme jointly prepared by NUCLEBRAS and KFA-Juelich. Brazil is planning to use elements of its own production in nuclear power plants to be erected within the German-Brazilian Industrial Agreement. As no material test reactor is available in Brazil it is expedient to irradiate samples of Brazilian production in Germany. Brazilian collaborators will participate in the preparation, execution and post-irradiation examination. In this way an optimum transfer of all information and results is assured. In the first phase, sample rods manufactured in Brazil are irradiated in the FRJ-2 test reactor in Juelich. These rods are assembled under clean conditions in the NUCLEBRAS research centres. The first Brazilian test rods showed excellent in-pile behaviour even under very high fuel rod capacity. In the second phase, fuel rods of original length manufactured and assembled in Brazil will be irradiated in German power plants, and, at the same time, additional irradiations of small samples will be carried out in test reactors. In the third phase, rod clusters and complete fuel elements will be manufactured in Brazil and irradiated in German power plants until target burn-up. All the necessary prerequisites have been fulfilled to meet the above requirements, i.e. mutual interest, good infrastructure maintained by both partners, qualified personnel and last but not least unbureaucratic and effective help by the coordinating offices of NUCLEBRAS and KFA

Low-enriched research reactor fuel: Post-Irradiation Examinations at SCK-CEN

International Nuclear Information System (INIS)

Van den Berghe, S.; Leenaers, A.

2007-01-01

Generally, research and test reactors are fuelled with fuel plates instead of pins. In most cases in the past, these plates consisted of high enriched (higher than 95 percent 235 U) UAl 3 powder mixed with a pure Al matrix (called the meat) in between two aluminium alloy plates (the cladding). These plates are then assembled in fuel elements of different designs to fit the needs of the various reactors. Since the 1970's, efforts have been going on to replace the high-enriched, low-density UAl 3 fuel with high-density, low enriched ( 235 U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched materials because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative and the Reduced Enrichment for Research and Test Reactors program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has been obtained with U 3 Si 2 fuel, which is currently used in many research reactors in the world. However, efforts to search for a better replacement have continued and are currently directed towards the U-Mo alloy fuel (7-10 weight percent Mo)

Characterization of irradiated fuel rods using pulsed eddy current techniques

International Nuclear Information System (INIS)

Martin, M.R.; Francis, W.C.

1975-11-01

A number of irradiated fuel rods and unfueled zircaloy cladding tubes (''water tubes'') were obtained from the Saxton reactor through arrangements with the Westinghouse Electric Corporation for use in subsequent irradiation effects and fuel behavior programs. A comprehensive nondestructive and corroborative destructive characterization program was undertaken on these fuel rods and tubes by ANC to provide baseline data on their characteristics prior to further testing and for comparison against post-post data. This report deals primarily with one portion of the NDT program performed remotely in the hot cells. The portion of interest in this paper is the pulsed eddy current inspection used in the nondestructive phase of the work. 6 references

Technology Implementation Plan: Irradiation Testing and Qualification for Nuclear Thermal Propulsion Fuel

Energy Technology Data Exchange (ETDEWEB)

Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rader, Jordan D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

This document is a notional technology implementation plan (TIP) for the development, testing, and qualification of a prototypic fuel element to support design and construction of a nuclear thermal propulsion (NTP) engine, specifically its pre-flight ground test. This TIP outlines a generic methodology for the progression from non-nuclear out-of-pile (OOP) testing through nuclear in-pile (IP) testing, at operational temperatures, flows, and specific powers, of an NTP fuel element in an existing test reactor. Subsequent post-irradiation examination (PIE) will occur in existing radiological facilities. Further, the methodology is intended to be nonspecific with respect to fuel types and irradiation or examination facilities. The goals of OOP and IP testing are to provide confidence in the operational performance of fuel system concepts and provide data to program leadership for system optimization and fuel down-selection. The test methodology, parameters, collected data, and analytical results from OOP, IP, and PIE will be documented for reference by the NTP operator and the appropriate regulatory and oversight authorities. Final full-scale integrated testing would be performed separately by the reactor operator as part of the preflight ground test.

Defect sizing of post-irradiated nuclear fuels using grayscale thresholding in their radiographic images

International Nuclear Information System (INIS)

Chaudhary, Usman Khurshid; Iqbal, Masood; Ahmad, Munir

2010-01-01

Quantification of different types of material defects in a number of reference standard post-irradiated nuclear fuel image samples have been carried out by virtue of developing a computer program that takes radiographic images of the fuel as input. The program is based on user adjustable grayscale thresholding in the regime of image segmentation whereby it selects and counts the pixels having graylevel values less than or equal to the computed threshold. It can size the defects due to chipping in nuclear fuel, cracks, voids, melting, deformation, inclusion of foreign materials, heavy isotope accumulation, non-uniformity, etc. The classes of fuel range from those of research and power reactors to fast breeders and from pellets to annular and vibro-compacted fuel. The program has been validated against ground truth realities of some locally fabricated metallic plates having drilled holes of known sizes simulated as defects in them in which the results indicate that it either correctly selects and quantifies at least 94% of the actual required regions of interest in a given image or it gives less than 8.1% false alarm rate. Also, the developed program is independent of image size.

Defect sizing of post-irradiated nuclear fuels using grayscale thresholding in their radiographic images

Energy Technology Data Exchange (ETDEWEB)

Chaudhary, Usman Khurshid, E-mail: ukhurshid@hotmail.co [Department of Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan); Iqbal, Masood, E-mail: masiqbal@hotmail.co [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad 45650 (Pakistan); Ahmad, Munir [Nondestructive Testing Group, Directorate of Technology, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad 45650 (Pakistan)

2010-10-15

Quantification of different types of material defects in a number of reference standard post-irradiated nuclear fuel image samples have been carried out by virtue of developing a computer program that takes radiographic images of the fuel as input. The program is based on user adjustable grayscale thresholding in the regime of image segmentation whereby it selects and counts the pixels having graylevel values less than or equal to the computed threshold. It can size the defects due to chipping in nuclear fuel, cracks, voids, melting, deformation, inclusion of foreign materials, heavy isotope accumulation, non-uniformity, etc. The classes of fuel range from those of research and power reactors to fast breeders and from pellets to annular and vibro-compacted fuel. The program has been validated against ground truth realities of some locally fabricated metallic plates having drilled holes of known sizes simulated as defects in them in which the results indicate that it either correctly selects and quantifies at least 94% of the actual required regions of interest in a given image or it gives less than 8.1% false alarm rate. Also, the developed program is independent of image size.

An experiment to examine the mechanistic behaviour of irradiated CANDU fuel stored under dry conditions

International Nuclear Information System (INIS)

Oldaker, I.E.; Crosthwaite, J.L.; Keltie, R.J.; Truss, K.J.

1979-01-01

A program has begun to use the Whiteshell Nuclear Research Establishment dry-storage canisters to store some selected CANDU irradiated fuel bundles in an 'easily retrievable basket.' The object of the experimental program is to study the long-term stability of the Zircaloy-sheathed UO 2 and UC fuel elements when stored in air. Bundles were loaded into a canister in October 1979 following detailed examination and removal of up to three complete elements from most bundles. These elements are currently being subjected to detailed destructive examinations, including metallography and scanning electron micrography, to fully characterize their pre-storage condition. After four years, and every five years thereafter, further elements will be examined similarly to study the effects of the storage environment on the stability of the Zircaloy sheathing, and on its continued ability to contain the fuel safely in an interim storage facility. (author)

Experimental Irradiations of Materials and Fuels in the BR2 Reactor: An Overview of Current Programmes

International Nuclear Information System (INIS)

Van Dyck, S.; Koonen, E.; Verwerft, M.; Wéber, M.

2013-01-01

The BR2 material test reactor offers a variety of experimental irradiation possibilities for testing of materials, fuels and instruments. The current paper gives an overview of the recent and ongoing programmes in order to illustrate the experimental potential of the reactor. Three domains of applications are reviewed: Irradiation of materials and fuels for pressurised water reactors (PWR); irradiation of materials for accelerator driven systems (ADS), cooled by liquid lead alloys; and irradiation of fuel for Material Test Reactors (MTR). For PWR relevant tests, a dedicated loop is available, providing a full simulation of the thermo hydraulic conditions of a PWR. ADS related tests require particular control of the irradiation environment and the necessary safety precautions in order to avoid 210 Po contamination. In-core mechanical testing of materials is done in comparison and complimentarily to post-irradiation examinations in order to assess flux related effects on the deformation behaviour of materials. (author)

CEA fuel pencil qualification under irradiation: from component conception to fuel assembly irradiation in a power reactor

International Nuclear Information System (INIS)

Marin, J.-F.; Pillet, Claude; Francois, Bernard; Morize, Pierre; Petitgrand, Sylvie; Atabek, R.-M.; Houdaille, Brigitte.

1981-06-01

Fabrication of fuel pins made of uranium oxide pellets and of a zircaloy 4 cladding is described. Irradiation experiment results are given. Thermomechanical behavior of the fuel pin in a power reactor is examined [fr

The French post irradiation examination database for the validation of depletion calculation tools

International Nuclear Information System (INIS)

Roque, Benedicte; Marimbeau, Pierre; Bioux, Philippe; Toubon, Herve; Daudin, Lucien

2003-01-01

This paper presents the experimental programmes conducted in France by the Commissariat a l'Energie Atomique (CEA) in order to validate spent fuel inventory calculations for core studies as well as fuel cycle studies. This large experimental programme was obtained in collaboration with our French partners, Electricite de France (EDF), FRAMATOME-ANP and COGEMA. The experimental data are based on chemical analysis measurements from fuel rod cuts irradiated in French reactors for PWR-UOx and MOx fuels, then dissolved in CEA laboratories, and from full assembly dissolutions at the COGEMA/La Hague reprocessing plants for UOx fuels. This enables us to cover a large range of UOx fuels with various enrichments in 235 U, 3.1% to 4.5%, associated with burnups from 10 GWd/t to 60 GWd/t. Recently, MOx fuels have also been investigated, with an initial Pu amount in the central zone of 5.6% and a maximum burnup of 45 GWd/t. Uranium, Plutonium, Americium, Curium isotopes and some fission products were analysed. Furthermore, Fission Products involved in Burn up Credit studies were measured. The experimental database contains also data for Boiling Water Reactor (BWR) with irradiated samples of BWR 9x9 and full BWR assemblies dissolutions. Furthermore some data exist for Fast Breeder Reactor (FBR) with small samples irradiated in the PHENIX reactor. An overview of ongoing programmes is also presented. (author)

Summary report on the fuel performance modeling of the AFC-2A, 2B irradiation experiments

Energy Technology Data Exchange (ETDEWEB)

Pavel G. Medvedev

2013-09-01

The primary objective of this work at the Idaho National Laboratory (INL) is to determine the fuel and cladding temperature history during irradiation of the AFC-2A, 2B transmutation metallic fuel alloy irradiation experiments containing transuranic and rare earth elements. Addition of the rare earth elements intends to simulate potential fission product carry-over from pyro-metallurgical reprocessing. Post irradiation examination of the AFC-2A, 2B rodlets revealed breaches in the rodlets and fuel melting which was attributed to the release of the fission gas into the helium gap between the rodlet cladding and the capsule which houses six individually encapsulated rodlets. This release is not anticipated during nominal operation of the AFC irradiation vehicle that features a double encapsulated design in which sodium bonded metallic fuel is separated from the ATR coolant by the cladding and the capsule walls. The modeling effort is focused on assessing effects of this unanticipated event on the fuel and cladding temperature with an objective to compare calculated results with the temperature limits of the fuel and the cladding.

Development of non-destructive examination system for irradiated fuel rods

International Nuclear Information System (INIS)

Sumerling, R.; Goldsmith, L.A.; Cross, M.T.; McKee, F.

1978-12-01

The development of non-destructive examination (NDE) system for irradiated fuel rods is described. The system is used for testing rods within a concrete cave and consists of three parts: a fully-automated fuel rod-drive machine, designed for easy maintenance; a series of plug-in NDE modules which fit into the central space provided in the machine, plus optical/TV viewing devices and gamma-scan equipment lined up on the rod; and on electronic control equipment situated outside the concrete shielding. The equipment is at present routinely used for viewing, eddy-current testing, gamma-scanning and diameter measurement of rods. The system is flexible in that additional modules can be added later as they are developed, since there is room for three modules of standard size (about 10cm x 10 cm x 3cm) in the machine or one large module taking the full space. New developments include the use of dual frequency eddy-current testing, which allows much greater discrimination against unwanted signals, and measurement of oxide thickness using a high frequency eddy-current probe. (author)

Computer-controlled gamma-ray scanner for irradiated reactor fuel

International Nuclear Information System (INIS)

Mandler, J.W.; Coates, R.A.; Killian, E.W.

1979-01-01

Gamma-ray scanning of irradiated fuel is an important nondestructive technique used in the thermal fuels behavior program currently under way at the Idaho National Engineering Laboratory. This paper is concerned with the computer-controlled isotopic gamma-ray-scanning system developed for postirradiation examination of fuel and includes a brief discussion of some scan results obtained from fuel rods irradiated in the Power-Burst Facility to illustrate gamma-ray spectrometry for this application. Both burnup profiles and information concerning fission-product migration in irradiated fuel are routinely obtained with the computer-controlled system

Irradiation of Superheater Test Fuel Elements in the Steam Loop of the R2 Reactor

Energy Technology Data Exchange (ETDEWEB)

Ravndal, F

1967-12-15

The design, fabrication, irradiation results, and post-irradiation examination for three superheater test fuel elements are described. During the spring of 1966 these clusters, each consisting of six fuel rods, were successfully exposed in the superheater loop No. 5 in the R2 reactor for a maximum of 24 days at a maximum outer cladding surface temperature of {approx} 650 deg C. During irradiation the linear heat rating of the rods was in the range 400-535 W/cm. The diameter of the UO{sub 2} pellets was 11.5 and 13.0 mm; the wall thickness of the 20/25 Nb and 20/35 cladding was in every case 0.4 mm. The diametrical gap between fuel and cladding was one of the main parameters and was chosen to be 0.05, 0.07 and 0.10 mm. These experiments, to be followed by one high cladding temperature irradiation ({approx} 750 deg C) and one long time irradiation ({approx} 6000 MWd/tU), were carried out to demonstrate the operational capability of short superheater test fuel rods at steady and transient operational environments for the Marviken superheater fuel elements and also to provide confirmation of design criteria for the same fuel elements.

Thermophysical instruments for non-destructive examination of tightness and internal gas pressure or irradiated power reactor fuel rods

International Nuclear Information System (INIS)

Pastoushin, V.V.; Novikov, A.Yu.; Bibilashvili, Yu.K.

1998-01-01

The developed thermophysical method and technical instruments for non-destructive leak-tightness and gas pressure inspection inside irradiated power reactor fuel rods and FAs under poolside and hot cell conditions are described. The method of gas pressure measuring based on the examination of parameters of thermal convection that aroused in gas volume of rod plenum by special technical instruments. The developed method and technique allows accurate value determination of not only one of the main critical rod parameters, namely total internal gas pressure, that forms rod mean life in the reactor core, but also the partial pressure of every main constituent of gaseous mixture inside irradiated fuel rod, that provides the feasibility of authentic and reliable leak-tightness detection. The described techniques were experimentally checked during the examination of all types power reactor fuel rods existing in Russia (WWER, BN, RBMK) and could form the basis for new technique development for non-destructive examination of PWR (and other) type rods and FAs having gas plenum filled with spring or another elements of design. (author)

In-pile irradiation test program and safety analysis report of the KAERI fuel for HANARO

Energy Technology Data Exchange (ETDEWEB)

Hwang, Wan; Ryu, Woo Suck; Byun, Taek Sang; Park, Jong Man; Lee, Byung Chul; Kim, Hack No; Park, Hee Tae; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-05-01

Localization of HANARO fuel has been carried out successfully, and design and fabrication technologies of the fuel are recently arrived the final stage of development. The performance of the fuel which has been fabricated in KAERI is confirmed through out-of-pile characterization, and the quality assurance procedure and assessment criteria are described. In order to verify the KAERI fuel, thus, in-pile irradiation test program of the KAERI fuel is scheduled in HANARO. This report summarizes the in-pile testing schedule, design documents of test rods and assemblies, fabrication history and out-of-pile characteristics of test rods, irradiation test condition and power history, post-irradiation examination scheme, linear power generation distribution, and safety analysis results. The design code for HANARO fuel is used to analyze the centerline temperature and swelling of the KAERI fuels. The results show that at 120 kW/m of linear power the maximum centerline temperature is 267 deg C which is much lower than the limitation temperature of 350 deg C, and that the swelling is 9.3 % at 95 at% lower than criterion of 20 %. Therefore, the KAERI fuels of this in-pile irradiation test is assessed to show good performance of integrity and safety in HANARO. 10 tabs., 7 figs., 3 refs. (Author).

Fission product phases in irradiated carbide fuels

International Nuclear Information System (INIS)

Ewart, F.T.; Sharpe, B.M.; Taylor, R.G.

1975-09-01

Oxide fuels have been widely adopted as 'first charge' fuels for demonstration fast reactors. However, because of the improved breeding characteristics, carbides are being investigated in a number of laboratories as possible advanced fuels. Irradiation experiments on uranium and mixed uranium-plutonium carbides have been widely reported but the instances where segregate phases have been found and subjected to electron probe analysis are relatively few. Several observations of such segregate phases have now been made over a period of time and these are collected together in this document. Some seven fuel pins have been examined. Two of the irradiations were in thermal materials testing reactors (MTR); the remainder were experimental assemblies of carbide gas bonded oxycarbide and sodium bonded oxycarbide in the Dounreay Fast Reactor (DFR). All fuel pins completed their irradiation without failure. (author)

Irradiation performance of AGR-1 high temperature reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Hunn, John D. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Ploger, Scott A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Morris, Robert N.; Baldwin, Charles A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Harp, Jason M.; Winston, Philip L. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Gerczak, Tyler J. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Rooyen, Isabella J. van [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Montgomery, Fred C.; Silva, Chinthaka M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States)

2016-09-15

Highlights: • Post-irradiation examination was performed on AGR-1 coated particle fuel. • Cesium release from the particles was very low in the absence of failed SiC layers. • Silver release was often substantial, and varied considerably with temperature. • Buffer and IPyC layers were found to play a key role in TRISO coating behavior. • Fission products palladium and silver were found in the SiC layer of particles. - Abstract: The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of {sup 110m}Ag from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10{sup −4} to 5 × 10{sup −4} for {sup 154}Eu and 8 × 10{sup −7} to 3 × 10{sup −5} for {sup 90}Sr. The average {sup 134}Cs fractional release from compacts was <3 × 10{sup −6} when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10{sup 5} in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving {sup 134}Cs fractional release in two capsules to approximately 10{sup −5}. Identification and characterization of these particles has provided unprecedented insight into

Advanced disassembling technique of irradiated driver fuel assembly for continuous irradiation of fuel pins

International Nuclear Information System (INIS)

Ichikawa, Shoichi; Haga, Hiroyuki; Katsuyama, Kozo; Maeda, Koji; Nishinoiri, Kenji

2012-01-01

It was necessary to carry out continuous irradiation tests in order to obtain the irradiation data of high burn-up fuel and high neutron dose material for FaCT (Fast Reactor Cycle Technology Development) project. There, the disassembling technique of an irradiated fuel assembly was advanced in order to realize further continuous irradiation tests. Although the conventional disassembling technique had been cutting a lower end-plug of a fuel pin needed to fix fuel pins to an irradiation vehicle, the advanced disassembling technique did not need cutting a lower end-plug. As a result, it was possible to supply many irradiated fuel pins to various continuous irradiation tests for FaCT project. (author)

Experimental and thermodynamic evaluation of the melting behavior of irradiated oxide fuels

International Nuclear Information System (INIS)

Adamson, M.G.; Aitken, E.A.; Caputi, R.W.

1985-01-01

Onset of melting is an important performance limit for irradiated UO 2 and UO 2 -based nuclear reactor fuels. Melting (solidus) temperatures are reasonably well known for starting fuel materials such as UO 2 and (U,PU)O 2 , however the influence of burnup on oxide fuel melting behavior continues to represent an area of considerable uncertainty. In this paper we report the results of a variety of melting temperature measurements on pseudo-binary fuel-fissia mixtures such as UO 2 -PUO 2 , UO 2 -CeO 2 , UO 2 -BaO, UO 2 -SrO, UO 2 -BaZrO 3 and UO 2 -SrZrO 3 . These measurements were performed using the thermal arrest technique on tungsten-encapsulated specimens. Several low melting eutectics, the existence of which had previously been inferred from post-irradiation examinations of high burnup mixed oxide fuels, were characterized in the course of the investigation. Also, an assessment of melting temperature changes in irradiated oxide fuels due to the production and incorporation of soluble oxidic fission products was performed by application of solution theory to the available pseudo-binary phase diagram data. The results of this assessment suggest that depression of oxide fuel solidus temperatures by dissolved fission products is substantially less than that indicated by earlier experimental studies. (orig.)

Nondestructive examination techniques on Candu fuel elements

International Nuclear Information System (INIS)

Gheorghe, G.; Man, I.

2013-01-01

During irradiation in nuclear reactor, fuel elements undergo dimensional and structural changes, and changes of surface conditions sheath as well, which can lead to damages and even loss of integrity. Visual examination and photography of Candu fuel elements are among the non-destructive examination techniques, next to dimensional measurements that include profiling (diameter, bending, camber) and length, sheath integrity control with eddy currents, measurement of the oxide layer thickness by eddy current techniques. Unirradiated Zircaloy-4 tubes were used for calibration purposes, whereas irradiated Zircaloy-4 tubes were actually subjected to visual inspection and dimensional measurements. We present results of measurements done by eddy current techniques on Zircaloy- 4 tubes, unirradiated, but oxidized in an autoclave prior to examinations. The purpose of these nondestructive examination techniques is to determine those parameters that characterize the behavior and performance of nuclear fuel operation. (authors)

Fission gas release during post irradiation annealing of large grain size fuels from Hinkley point B

International Nuclear Information System (INIS)

Killeen, J.C.

1997-01-01

A series of post-irradiation anneals has been carried out on fuel taken from an experimental stringer from Hinkley Point B AGR. The stringer was part of an experimental programme in the reactor to study the effect of large grain size fuel. Three differing fuel types were present in separate pins in the stringer. One variant of large grain size fuel had been prepared by using an MgO dopant during fuel manufactured, a second by high temperature sintering of standard fuel and the third was a reference, 12μm grain size fuel. Both large grain size variants had similar grain sizes around 35μm. The present experiments took fuel samples from highly rated pins from the stringer with local burn-up in excess of 25GWd/tU and annealed these to temperature of up to 1535 deg. C under reducing conditions to allow a comparison of fission gas behaviour at high release levels. The results demonstrate the beneficial effect of large grain size on release rate of 85 Kr following interlinkage. At low temperatures and release rates there was no difference between the fuel types, but at temperatures in excess of 1400 deg. C the release rate was found to be inversely dependent on the fuel grain size. The experiments showed some differences between the doped and undoped large grains size fuel in that the former became interlinked at a lower temperature, releasing fission gas at an increased rate at this temperature. At higher temperatures the grain size effect was dominant. The temperature dependence for fission gas release was determined over a narrow range of temperature and found to be similar for all three types and for both pre-interlinkage and post-interlinkage releases, the difference between the release rates is then seen to be controlled by grain size. (author). 4 refs, 7 figs, 3 tabs

Fission gas release during post irradiation annealing of large grain size fuels from Hinkley point B

Energy Technology Data Exchange (ETDEWEB)

Killeen, J C [Nuclear Electric plc, Barnwood (United Kingdom)

1997-08-01

A series of post-irradiation anneals has been carried out on fuel taken from an experimental stringer from Hinkley Point B AGR. The stringer was part of an experimental programme in the reactor to study the effect of large grain size fuel. Three differing fuel types were present in separate pins in the stringer. One variant of large grain size fuel had been prepared by using an MgO dopant during fuel manufactured, a second by high temperature sintering of standard fuel and the third was a reference, 12{mu}m grain size fuel. Both large grain size variants had similar grain sizes around 35{mu}m. The present experiments took fuel samples from highly rated pins from the stringer with local burn-up in excess of 25GWd/tU and annealed these to temperature of up to 1535 deg. C under reducing conditions to allow a comparison of fission gas behaviour at high release levels. The results demonstrate the beneficial effect of large grain size on release rate of {sup 85}Kr following interlinkage. At low temperatures and release rates there was no difference between the fuel types, but at temperatures in excess of 1400 deg. C the release rate was found to be inversely dependent on the fuel grain size. The experiments showed some differences between the doped and undoped large grains size fuel in that the former became interlinked at a lower temperature, releasing fission gas at an increased rate at this temperature. At higher temperatures the grain size effect was dominant. The temperature dependence for fission gas release was determined over a narrow range of temperature and found to be similar for all three types and for both pre-interlinkage and post-interlinkage releases, the difference between the release rates is then seen to be controlled by grain size. (author). 4 refs, 7 figs, 3 tabs.

Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

1993-02-01

This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

Irradiation testing of coated particle fuel at Hanaro

International Nuclear Information System (INIS)

Goo Kim, Bong; Sung Cho, Moo; Kim, Yong Wan

2014-01-01

TRISO-coated particle fuel is developing to support development of VHTR in Korea. From August 2013, the first irradiation testing of coated particle fuel was begun to demonstrate and qualify TRISO fuel for use in VHTR in the HANARO at KAERI. This experiment is currently undergoing under the atmosphere of a mixed inert gas without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one contains nine fuel compacts and the other five compacts and eight graphite specimens. Each compact has 263 coated particles. After a peak burn-up of about 4 at% and a peak fast neutron fluence of about 1.7 x 10 21 n/cm 2 , PIE will be carried out at KAERI's Irradiated Material Examination Facility. This paper is described characteristics of coated particle fuel, the design of test rod and irradiation device for coated particle fuel, discusses the technical results for irradiation testing at HANARO. (authors)

Irradiation of Argentine (U,Pu)O2 MOX fuels. Post-irradiation results and experimental analysis with the BACO code

International Nuclear Information System (INIS)

Marino, A.C.; Perez, E.; Adelfang, P.

1996-01-01

The irradiation of the first Argentine prototypes of pressurized heavy water reactor (PHWR) (U,Pu)O 2 MOX fuels began in 1986. These experiments were carried out in the High Flux Reactor (HFR)-Petten, Holland. The rods were prepared and controlled in the C NEA's α Facility. The postirradiation examinations were performed in the Kernforschungszentrum, Karlsruhe, Germany and in the Joint Research Center (JRC), Petten. The first rod has been used for destructive pre-irradiation analysis. The second one as a pathfinder to adjust systems in the HFR. Two additional rods including iodine doped pellets were intended to simulate 15000 MWd/T(M) burnup. The remaining two rods were irradiated until 15000 MWd/T(M). One of them underwent a final ramp with the aim of verifying fabrication processes and studying the behaviour under power transients. BACO (BArra COmbustible) code was used to define the power histories and to analyse the experiments. This paper presents a description of the different experiments and a comparison between the results of the postirradiation examinations and the BACO outputs. (orig.)

Irradiation of Argentine (U,Pu)O 2 MOX fuels. Post-irradiation results and experimental analysis with the BACO code

Science.gov (United States)

Marino, Armando Carlos; Pérez, Edmundo; Adelfang, Pablo

1996-04-01

The irradiation of the first Argentine prototypes of pressurized heavy water reactor (PHWR) (U,Pu)O 2 MOX fuels began in 1986. These experiments were carried out in the High Flux Reactor (HFR)-Petten, Holland. The rods were prepared and controlled in the CNEA's α Facility. The postirradiation examinations were performed in the Kernforschungszentrum, Karlsruhe, Germany and in the Joint Research Center (JRC), Petten. The first rod has been used for destructive pre-irradiation analysis. The second one as a pathfinder to adjust systems in the HFR. Two additional rods including iodine doped pellets were intended to simulate 15 000 MWd/T(M) burnup. The remaining two rods were irradiated until 15 000 MWd/T(M). One of them underwent a final ramp with the aim of verifying fabrication processes and studying the behaviour under power transients. BACO (BArra COmbustible) code was used to define the power histories and to analyse the experiments. This paper presents a description of the different experiments and a comparison between the results of the postirradiation examinations and the BACO outputs.

Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

1992-01-01

This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

SP-100 Fuel Pin Performance: Results from Irradiation Testing

Science.gov (United States)

Makenas, Bruce J.; Paxton, Dean M.; Vaidyanathan, Swaminathan; Marietta, Martin; Hoth, Carl W.

1994-07-01

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 pins 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.

Information for irradiation and post-irradiation of the silicide fuel element prototype P-07

International Nuclear Information System (INIS)

Abbate, Maximo J.; Sbaffoni, Maria M.

2003-01-01

Included in the 'Silicides' Project, developed by the Nuclear Fuels Department of the National Atomic Energy Commission (CNEA), it is foreseen the qualification of this type of fuel for research reactors in order to be used in the Argentine RA-3 reactor and to confirm the CNEA as an international supplier. The paper presents basic information on several parameters corresponding to the new silicide prototype, called P-07, to be taken into account for its irradiation, postirradiation and qualification. (author)

Post-irradiation examinations and high-temperature tests on undoped large-grain UO{sub 2} discs

Energy Technology Data Exchange (ETDEWEB)

Noirot, J., E-mail: jean.noirot@cea.fr [CEA, DEN, DEC, Cadarache, F-13108 St. Paul Lez Durance (France); Pontillon, Y. [CEA, DEN, DEC, Cadarache, F-13108 St. Paul Lez Durance (France); Yagnik, S. [EPRI, P.O. Box 10412, Palo Alto, CA 94303-0813 (United States); Turnbull, J.A. [Independent Consultant (United Kingdom)

2015-07-15

Within the Nuclear Fuel Industry Research (NFIR) programme, several fuel variants –in the form of thin circular discs – were irradiated in the Halden Boiling Water Reactor (HBWR) at burn-ups up to ∼100 GWd/t{sub HM}. The design of the fuel assembly was similar to that used in other HBWR programmes: the assembly contained several rods with fuel discs sandwiched between Mo discs, which limited temperature differences within each fuel disc. One such variant was made of large-grain UO{sub 2} discs (3D grain size = ∼45 μm) which were subjected to three burn-ups: 42, 72 and 96 GWd/t{sub HM}. Detailed characterizations of some of these irradiated large-grain UO{sub 2} discs were performed in the CEA Cadarache LECA-STAR hot laboratory. The techniques used included electron probe microanalysis (EPMA), scanning electron microscopy (SEM) and secondary ion mass spectrometry (SIMS). Comparisons were then carried out with more standard grain size UO{sub 2} discs irradiated under the same conditions. Examination of the high burn-up large-grain UO{sub 2} discs revealed the limited formation of a high burn-up structure (HBS) when compared with the standard-grain UO{sub 2} discs at similar burn-up. High burn-up discs were submitted to temperature transients up to 1200 °C in the heating test device called Merarg at a relatively low temperature ramp rate (0.2 °C/s). In addition to the total gas release during these tests, the release peaks throughout the temperature ramp were monitored. Tests at 1600 °C were also conducted on the 42 GWd/t{sub HM} discs. The fuels were then characterized with the same microanalysis techniques as those used before the tests, to investigate the effects of these tests on the fuel’s microstructure and on the fission gas behaviour. This paper outlines the high resistance of this fuel to gas precipitation at high temperature and to HBS formation at high burn-up. It also shows the similarity of the positions, within the grains, where HBS forms

Metallographic examination in irradiated materials examination facility

Energy Technology Data Exchange (ETDEWEB)

Choo, Yong Sun; Lee, Key Soon; Park, Dae Gyu; Ahn, Sang Bok; Yoo, Byoung Ok

1998-01-01

It is very important to have equipment of metallographic examination in hot-cell to observe the micro-structure of nuclear fuels and materials irradiated at nuclear power and/or research reactor. Those equipment should be operated by master-slave manipulators, so they are designed, manufactured and modified to make exercise easy and no trouble. The metallographic examination equipment and techniques as well as its operation procedure are described, so an operator can practice the metallography in hot-cell. (author). 5 refs., 7 tabs., 21 figs.

Irradiation of defected SAP clad UO2 fuel in the X-7 organic loop

International Nuclear Information System (INIS)

Robertson, R.F.S.; Cracknell, A.G.; MacDonald, R.D.

1961-10-01

This report describes an experiment designed to test the behaviour under irradiation of a UO 2 fuel specimen clad in a defected SAP sheath and cooled by recirculating organic liquid. The specimen containing the defect was irradiated in the X-7 loop in the NRX reactor from the 25th of November until the 13th of December 1960. Up to the 13th of December the behaviour was analogous to that seen with defected UO 2 specimens clad in zircaloy which were irradiated in water loops. Reactor power transients resulted in peaking of gamma ray activities in the loop, but on steady operation these activities tended to fall to a steady state level, Over this period the pressure drop across the fuel increased by a factor of two, the increases occurring after reactor shut downs and start ups. On 13th December the pressure drop increased rapidly, after a reactor shut down and start up, to over five times its original value and the activities in the loop rose to a high level. The specimen was removed and examination showed that the sheath was very badly split and that the volume between the fuel and the sheath was filled with a hard black organic substance. This report gives full details of the irradiation and of the post -irradiation examination. Correlation of the observed phenomenon is attempted and a preliminary assessment of the problems which would be associated with defect fuel in an organic reactor is given. (author)

DUPIC fuel fabrication using spent PWR fuels at KAERI

Energy Technology Data Exchange (ETDEWEB)

Kim, Ho Dong; Yang, Myung Seung; Ko, Won Il and others

2000-12-01

This document contains DUPIC fuel cycle R and D activities to be carried out for 5 years beyond the scope described in the report KAERI/AR-510/98, which was attached to Joint Determination for Post-Irradiation Examination of irradiated nuclear fuel, by MOST and US Embassy in Korea, signed on April 8, 1999. This document is purposely prepared as early as possible to have ample time to review that the over-all DUPIC activities are within the scope and contents in compliance to Article 8(C) of ROK-U.S. cooperation agreement, and also maintain the current normal DUPIC project without interruption. Manufacturing Program of DUPIC Fuel in DFDF and Post Irradiation Examination of DUPIC Fuel are described in Chapter I and Chapter II, respectively. In Chapter III, safeguarding procedures in DFDF and on-going R and D on DUPIC safeguards such as development of nuclear material accounting system and development of containment/surveillance system are described in details.

An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Evans, Louise G [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, S. J. [Los Alamos National Laboratory; Menlove, H. O. [Los Alamos National Laboratory; Schear, M. A. [Los Alamos National Laboratory; Worrall, Andrew [U.K. NNL

2011-01-13

Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/ or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

Irradiation performance of experimental fast reactor 'JOYO' MK-1 driver fuel assemblies

International Nuclear Information System (INIS)

Itaki, Toshiyuki; Kono, Keiichi; Tachi, Hirokatsu; Yamanouchi, Sadamu; Yuhara, Shunichi; Shibahara, Itaru

1985-01-01

The experimental fast reactor ''JOYO'' completed it's breeder core (MK-I) operation in January 1982. The MK-I driver fuel assemblies were removed from the core sequencially in order of burnup increase and have been under postirradiation examination (PIE). The PIE has almost been completed for 30 assemblies including the highest burnup assemblies of 48,000 MWD/MTM. It has been confirmed that all fuel assemblies have exhibited satisfactory performance without detrimental assembly deformation or without any indications of fuel pin breach. The irradiation conditions of the MK-I core were somewhat more moderate than those conditions envisioned for prototypic reactor. However the results of the examination revealed the typical irradiation behavior of LMFBR fuels, although such characteristics were benign as compared with those anticipated in high burnup fuels. Systematic performance data have been accumulated through the fuel fabrication, irradiation and postirradiation examination processes. Based on these data, the MK-I fuel designing and fabrication techniques were totally confirmed. This technical experience and the associated insight into irradiation behavior have established a milestone to the next step of fast reactor fuel development. (author)

Metallographic analysis of irradiated RERTR-3 fuel test specimens

International Nuclear Information System (INIS)

Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

2000-01-01

The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date

Examinations of the irradiation behaviour of U3Si2 test fuel plates with low enrichment

International Nuclear Information System (INIS)

Muellauer, J.

1989-01-01

Five low-enriched (19.7% 235 U), high-density (4.7 gU/cm/ 3 ) U 3 Si 2 -test fuel plates (miniplates) with different fine grain contents have been qualified under irradiation. During the course of irradiation up to burnup of 63% 235 U depletion, no released fractions of gaseous or solid fission products from the fuel plate to the rig coolant were detected. The measured swelling rate of the fuel zone (meat) is less than 0.45% ΔV/10 20 fissions/cm 3 the blister-threshold temperature of the fuel plates is above 520 0 C. The favourable irradiation behavior of the U 3 Si 2 fuel plates was not influenced by using higher amounts of fine grained particles (40% [de

PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2

Energy Technology Data Exchange (ETDEWEB)

Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Morris, Robert Noel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Baldwin, Charles A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-04-01

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). This work is sponsored by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Advanced Reactor Technologies (ART) Office under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program. The AGR-1 experiment was the first in a series of TRISO fuel irradiation tests initiated in 2006. The AGR-1 TRISO particles and fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 using laboratory-scale equipment and irradiated for 3 years in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. Post-irradiation examination was performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE conducted at ORNL on loose particles extracted from irradiated AGR-1 Compact 4-4-2.

Improving thermal model prediction through statistical analysis of irradiation and post-irradiation data from AGR experiments

International Nuclear Information System (INIS)

Pham, Binh T.; Hawkes, Grant L.; Einerson, Jeffrey J.

2014-01-01

As part of the High Temperature Reactors (HTR) R and D program, a series of irradiation tests, designated as Advanced Gas-cooled Reactor (AGR), have been defined to support development and qualification of fuel design, fabrication process, and fuel performance under normal operation and accident conditions. The AGR tests employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule and instrumented with thermocouples (TC) embedded in graphite blocks enabling temperature control. While not possible to obtain by direct measurements in the tests, crucial fuel conditions (e.g., temperature, neutron fast fluence, and burnup) are calculated using core physics and thermal modeling codes. This paper is focused on AGR test fuel temperature predicted by the ABAQUS code's finite element-based thermal models. The work follows up on a previous study, in which several statistical analysis methods were adapted, implemented in the NGNP Data Management and Analysis System (NDMAS), and applied for qualification of AGR-1 thermocouple data. Abnormal trends in measured data revealed by the statistical analysis are traced to either measuring instrument deterioration or physical mechanisms in capsules that may have shifted the system thermal response. The main thrust of this work is to exploit the variety of data obtained in irradiation and post-irradiation examination (PIE) for assessment of modeling assumptions. As an example, the uneven reduction of the control gas gap in Capsule 5 found in the capsule metrology measurements in PIE helps identify mechanisms other than TC drift causing the decrease in TC readings. This suggests a more physics-based modification of the thermal model that leads to a better fit with experimental data, thus reducing model uncertainty and increasing confidence in the calculated fuel temperatures of the AGR-1 test

Improving thermal model prediction through statistical analysis of irradiation and post-irradiation data from AGR experiments

Energy Technology Data Exchange (ETDEWEB)

Pham, Binh T., E-mail: Binh.Pham@inl.gov [Human Factor, Controls and Statistics Department, Nuclear Science and Technology, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Hawkes, Grant L. [Thermal Science and Safety Analysis Department, Nuclear Science and Technology, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Einerson, Jeffrey J. [Human Factor, Controls and Statistics Department, Nuclear Science and Technology, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

2014-05-01

As part of the High Temperature Reactors (HTR) R and D program, a series of irradiation tests, designated as Advanced Gas-cooled Reactor (AGR), have been defined to support development and qualification of fuel design, fabrication process, and fuel performance under normal operation and accident conditions. The AGR tests employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule and instrumented with thermocouples (TC) embedded in graphite blocks enabling temperature control. While not possible to obtain by direct measurements in the tests, crucial fuel conditions (e.g., temperature, neutron fast fluence, and burnup) are calculated using core physics and thermal modeling codes. This paper is focused on AGR test fuel temperature predicted by the ABAQUS code's finite element-based thermal models. The work follows up on a previous study, in which several statistical analysis methods were adapted, implemented in the NGNP Data Management and Analysis System (NDMAS), and applied for qualification of AGR-1 thermocouple data. Abnormal trends in measured data revealed by the statistical analysis are traced to either measuring instrument deterioration or physical mechanisms in capsules that may have shifted the system thermal response. The main thrust of this work is to exploit the variety of data obtained in irradiation and post-irradiation examination (PIE) for assessment of modeling assumptions. As an example, the uneven reduction of the control gas gap in Capsule 5 found in the capsule metrology measurements in PIE helps identify mechanisms other than TC drift causing the decrease in TC readings. This suggests a more physics-based modification of the thermal model that leads to a better fit with experimental data, thus reducing model uncertainty and increasing confidence in the calculated fuel temperatures of the AGR-1 test.

The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

International Nuclear Information System (INIS)

Grover, S. Blaine

2009-01-01

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

Irradiation performance of metallic fuels

International Nuclear Information System (INIS)

Pahl, R.G.; Lahm, C.E.; Porter, D.L.; Batte, G.L.; Hofman, G.L.

1989-01-01

Argonne National Laboratory has been working for the past five years to develop and demonstrate the Integral Fast Reactor (IFR) concept. The concept involves a closed system for fast-reactor power generation and on-site fuel reprocessing, both designed specifically around the use of metallic fuel. The Experimental Breeder Reactor-II (EBR-II) has used metallic fuel for all of its 25-year life. In 1985, tests were begun to examine the irradiation performance of advanced-design metallic fuel systems based on U-Zr or U-Pu-Zr fuels. These tests have demonstrated the viable performance of these fuel systems to high burnup. The initial testing program will be described in this paper. 2 figs

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

Energy Technology Data Exchange (ETDEWEB)

Rest, J. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)], E-mail: jrest@anl.gov; Hofman, G.L.; Kim, Yeon Soo [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

2009-04-15

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than {approx}7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

Science.gov (United States)

Rest, J.; Hofman, G. L.; Kim, Yeon Soo

2009-04-01

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

Post-test examination of the VVER-1000 fuel rod bundle CORA-W2

International Nuclear Information System (INIS)

Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.; Schanz, G.; Hagen, S.; Sepold, L.

1995-01-01

The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B 4 C absorber rod and the stainless steel grid spacers on the 'low-temperature' bundle damage initiation and progression. The B 4 C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occured. (orig./HP)

Post-test examination of the VVER-1000 fuel rod bundle CORA-W2

Energy Technology Data Exchange (ETDEWEB)

Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.; Schanz, G.; Hagen, S.; Sepold, L.

1995-08-01

The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B{sub 4}C absorber rod and the stainless steel grid spacers on the `low-temperature` bundle damage initiation and progression. The B{sub 4}C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occured. (orig./HP)

Postirradiation examination of Kori-1 nuclear power plant fuels

International Nuclear Information System (INIS)

Ro, S.G.; Kim, E.K.; Lee, K.S.; Min, D.K.

1994-01-01

Full size fuels discharged from Kori-1 PWR nuclear power plant have been subjected to postirradiation examination. The fuels under investigation were irradiated for one- to four-reactor cycles. Nondestructive examination and dismantling of the fuel assemblies have been conducted in the pool of the postirradiation examination facility (PIEF) of Korea Atomic Energy Research Institue. Subsequently nondestructive and destructive examinations of fuel rods have been performed in the hot cells of the PIEF. An evaluation of fuel burnup behaviors was based on the postirradiation examination data and the nominal design values. The results did not show any evidence of abnormalities in the fuel integrity. (orig.)

Postirradiation examination of Kori-1 nuclear power plant fuels

Science.gov (United States)

Seung-Gy, Ro; Eun-Ka, Kim; Key-Soon, Lee; Duck-Kee, Min

1994-05-01

Full size fuels discharged from Kori-1 PWR nuclear power plant have been subjected to postirradiation examination. The fuels under investigation were irradiated for one- to four-reactor cycles. Nondestructive examination and dismantling of the fuel assemblies have been conducted in the pool of the postirradiation examination facility (PIEF) of Korea Atomic Energy Research Institute. Subsequently nondestructive and destructive examinations of fuel rods have been performed in the hot cells of the PIEF. An evaluation of fuel burnup behaviors was based on the postirradiation examination data and the nominal design values. The results did not show any evidence of abnormalities in the fuel integrity.

Experimental data report for test TS-3 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo; Sobajima, Makoto.

1993-09-01

This report presents experimental data for Test TS-3 which was the third test in a series of Reactivity Initiated Accident (RIA) tests using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in September, 1990. Test fuel rod used in the Test TS-3 was a short-sized BWR (7 x 7) type rod which was re-fabricated from a commercial rod irradiated in the Tsuruga Unit 1 power reactor of Japan Atomic Power Co. The fuel had an initial enrichment of 2.79 % and a burnup of 26 Gwd/tU. A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 94 ± 4 cal/g · fuel (88 ± 4 cal/g · fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and results of pre-pulse and post-pulse irradiation examinations are described in this report. (author)

Irradiation experiments of 3rd, 4th and 5th fuel assemblies by an in-pile gas loop, OGL-1

International Nuclear Information System (INIS)

Fukuda, Kousaku; Kobayashi, Fumiaki; Hayashi, Kimio; Minato, Kazuo; Kikuchi, Teruo; Adachi, Mamoru; Iwamoto, Kazumi; Ikawa, Katsuichi; Itami, Hiroharu.

1986-07-01

Three irradiation experiments for 3rd, 4th and 5th fuel assemblies which had been composed of VHTR reference coated particle fuels and graphite components were carried out by an in-pile gas loop, OGL-1 during 1979 and 1982. The main purposes of these experiments were to study on bowing of the fuel rod by irradiation for the 3rd fuel assembly, to study on fuel behavior under relatively low burnup irradiation for the 4th fuel assembly, and to study on fuel behavior up to full burnup of VHTR design for the 5th fuel assembly. For understanding in-pile fuel behavior, fractional releases of fission gases from each fuel assembly were estimated by measuring the fission gas concentrations in the primary loop of OGL-1. The post-irradiation examination (PIE) was carried out extensively on the fuel block, the fuel rods and the fuel compacts in Tokai Hot Laboratory. Also, made were the measurements of metallic fission product distributions in the fuel assemblies and the fuel rods. The results in these experiments were given as follows ; bowing of the fuel rod in the 3rd fuel assembly was 0.7 mm, but integrity of the rod was kept under irradiation. Fractional release of the fission gas from the 4th fuel assembly remained in the order of 10 -7 during irradiation, suggesting that the fuel performance was excellent. The fractional release from the 5th fuel assembly, on the other hand, was in the order of 10 -5 which was the same level in the VHTR design. (author)

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

Energy Technology Data Exchange (ETDEWEB)

Davenport, Michael; Petti, D. A.; Palmer, Joe

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

Dry storage of irradiated nuclear fuel

International Nuclear Information System (INIS)

Tolmie, R.D.

1983-01-01

In transferring radioactive material between the preparation and clean chambers of a dry storage complex, irradiated nuclear fuel is posted from the preparation chamber to a sealable canister supported in a closable bucket in the clean chamber, or a contaminated sealed canister is posted from a closed bucket in the clean chamber into the preparation chamber by using a facility comprising two coaxial tubes constituting a closable orifice between the two chambers, the tubes providing sealing means for the bucket, and masking means for the bucket and canister closures together with means for withdrawing the closures into the preparation chamber. (author)

Post-irradiation diarrhea

International Nuclear Information System (INIS)

Meerwaldt, J.H.

1984-01-01

In radiotherapy of pelvic cancers, the X-ray dose to be delivered to the tumour is limited by the tolerance of healthy surrounding tissue. In recent years, a number of serious complications of irradiation of pelvic organs were encountered. Modern radiotherapy necessitates the acceptance of a calculated risk of complications in order to achieve a better cure rate. To calculate these risks, one has to know the radiation dose-effect relationship of normal tissues. Of the normal tissues most at risk when treating pelvic tumours only the bowel is studied. In the literature regarding post-irradiation bowel complications, severe and mild complications are often mixed. In the present investigation the author concentrated on the group of patients with relatively mild symptoms. He studied the incidence and course of post-irradiation diarrhea in 196 patients treated for carcinoma of the uterine cervix or endometrium. The aims of the present study were: 1) to determine the incidence, course and prognostic significance of post-irradiation diarrhea; 2) to assess the influence of radiotherapy factors; 3) to study the relation of bile acid metabolism to post-irradiation diarrhea; 4) to investigate whether local factors (reservoir function) were primarily responsible. (Auth.)

Techniques and devices developed by the CEA for hot cell and in-situ examinations of PWR components and PWR fuel assembliess after irradiation

International Nuclear Information System (INIS)

Van Craeynest, J.C.; Leseur, A.; Lhermenier, A.; Cytermann, R.

1981-11-01

Within the framework of the electro-nuclear development of the PWR system, the CEA has provided itself with facilities for developing techniques for analyzing assemblies, pins and fuels. These are examinations and tests on irradiated heads and assemblies with the aid of the Fuel Examination Module (FEM), of machining of assemblies and examinations in the Celimene hot laboratory or detailed examinations and analyses on fuel elements using eddy currents, the electronic microprobe and the Fisher ''permeascope'' which enables the outline of the oxide coat present on the cladding to be followed [fr

Transport experience of NH-25 spent fuel shipping cask for post irradiation examination

International Nuclear Information System (INIS)

Mori, Ryuji

1982-01-01

Since the Japan Atomic Energy Research Institute and Nippon Nuclear Fuel Development Co. hot laboratories are located far off from the port which can handle spent fuel shipping casks, it is necessary to use a trailer-mounted cask which can be transported by public roads, bridges and intersections for the transportation of spent fuel specimens to these hot laboratories. Model NH-25 shipping cask was designed, manufactured and oualification tested to meet Japanese regulations and was officially registered as a BM type cask. The NH-25 cask accomodates two BWR fuel assemblies, one PWR assembly or one ATR fuel assembly using interchangeable inner containers. The cask weight is 29.2 t. The cask has three concentric stainless steel shells. Gamma shielding is lead cast between the inner shell and the intermediate shell. Neutro n shielding consists of ethylene-glycol-aqueous solution layer formed between the intermediate shell and the outer shell. The NH-25 cask now has been in operation for 2.5 yr. It was used for the transportation of spent fuel assemblies from six LWR power plants to the port on shipping cask carrier ''Hinouramaru'' on the sea, as well as from the port to the hot laboratory on a trailer. The capability of safe handling and transporting of spent fuel assemblies has been well demonstrated. (author)

Performance of Bruce natural UO2 fuel irradiated to extended burnups

International Nuclear Information System (INIS)

Zhou, Y.N.; Floyd, M.R.; Ryz, M.A.

1995-11-01

Bruce-type bundles XY, AAH and GF were successfully irradiated in the NRU reactor at Chalk River Laboratories to outer-element burnups of 570-900 MWh/kgU. These bundles were of the Bruce Nuclear Generating Station (NGS)-A 'first-charge' design that contained gas plenums in the outer elements. The maximum outer-element linear powers were 33-37 kW/m. Post-irradiation examination of these bundles confirmed that all the elements were intact. Bundles XY and AAH, irradiated to outer-element burnups of 570-700 MWh/kgU, experienced low fission-gas release (FGR) ( 500 MWh/kgU (equivalent to bundle-average 450 MWh/kgU) when maximum outer-element linear powers are > 50 kW/m. The analysis in this paper suggests that CANDU 37-element fuel can be successfully irradiated (low-FGR/defect-free) to burnups of at least 700 MWh/kgU, provided maximum power do not exceed 40 kW/m. (author). 5 refs., 1 tab., 8 figs

The Defect Inspection on the Irradiated Fuel Rod by Eddy Current Test

International Nuclear Information System (INIS)

Koo, D. S.; Park, Y. K.; Kim, E. K.

1996-01-01

The eddy current test(ECT) probe of differential encircling coil type was designed and fabricated, and the optimum condition of ECT was derived for the examination of the irradiated fuel rod. The correlation between ECT test frequency and phase and amplitude was derived by performing the test of the standard rig that includes inner notches, outer notches and through-holes. The defect of through-hole was predicted by ECT at the G33-N2 fuel rod irradiated in the Kori-1 nuclear power reactor. The metallographic examination on the G33-N2 fuel rod was Performed at the defect location predicted by ECT. The result of metallographic examination for the G33-N2 fuel rod was in good agreement with that of ECT. This proves that the evaluation for integrity of irradiated fuel rod by ECT is reliable

Neutron Focusing Mirrors for Neutron Radiography of Irradiated Nuclear Fuel at Idaho National Laboratory

Science.gov (United States)

Rai, Durgesh K.; Wu, Huarui; Abir, Muhammad; Giglio, Jeffrey; Khaykovich, Boris

Post irradiation examination (PIE) of samples irradiated in nuclear reactors is a challenging but necessary task for the development on novel nuclear power reactors. Idaho National Laboratory (INL) has neutron radiography capabilities, which are especially useful for the PIE of irradiated nuclear fuel. These capabilities are limited due to the extremely high gamma-ray radiation from the irradiated fuel, which precludes the use of standard digital detectors, in turn limiting the ability to do tomography and driving the cost of the measurements. In addition, the small 250 kW Neutron Radiography Reactor (NRAD) provides a relatively weak neutron flux, which leads to low signal-to-noise ratio. In this work, we develop neutron focusing optics suitable for the installation at NRAD. The optics would separate the sample and the detector, potentially allowing for the use of digital radiography detectors, and would provide significant intensity enhancement as well. The optics consist of several coaxial nested Wolter mirrors and is suited for polychromatic thermal neutron radiation. Laboratory Directed Research and Development program of Idaho National Laboratory.

Post irradiation examination of garter springs from Indian PHWRs

International Nuclear Information System (INIS)

Dubey, J.S.; Shah, Priti Kotak; Mishra, Prerna; Singh, H.N.; Alur, V.D.; Kumar, Ashwini; Bhandekar, A.; Pandit, K.M.; Anantharaman, S.

2013-12-01

Irradiated Zr-2.5Nb-0.5Cu garter springs, belonging to Indian Pressurised Heavy Water Reactors, which had experienced 8 to 10 Effective Full Power Years of operation were subjected to visual, dimensional, chemical, metallographic examination and relevant mechanical tests. Methodology of the tests conducted and results are presented. The digital photographs were used to measure the inner and outer circumferences by image processing. The hydrogen (H) content in the spring coils were measured using Differential Scanning Calorimetry (DSC). In the stretch test, all the irradiated GSs were found to require an additional load, as compared to unirradiated GS, to produce a given amount of residual extension which indicated that the irradiated GSs had undergone significant irradiation hardening. The crush test results showed that the minimum load required to crush the coil or cause a sudden sideways shift in the grips was higher than 400 N/coil, much higher than the design load. The test results indicated that the irradiated GS, after 10 EFPY of operation, have adequate strength and ductility to continue to meet the design intent. Mechanical tests were carried out on irradiated girdle wires taken out of the loose fit garter springs (GS) from (NAPS-1, ∼ 8.5 EFPY) and tight fit garter spring from KAPS-2 (∼ 8.0 EFPY) PHWRs. Tensile tests on the irradiated girdle wires, showed irradiation hardening in the material and reduction in ductility. The irradiated girdle wires have around 4 to 5% residual ductility level against the 15% ductility of unirradiated wire. The fracture surfaces of the irradiated as well as the un-irradiated girdle wires were observed in SEM. (author)

Image analysis for remote examination of fuel pins

International Nuclear Information System (INIS)

Cook, J.H.; Nayak, U.P.

1982-01-01

An image analysis system operating in the Wing 9 Hot Cell Facility at Los Alamos National Laboratory provides quantitative microstructural analyses of irradiated fuels and materials. With this system, fewer photomicrographs are required during postirradiation microstructural examination and data are available for analysis much faster. The system has been used successfully to examine Westinghouse Advanced Reactors Division experimental fuel pins

Results of post-irradiation examination of WWER fuel assembly structural components made of E110 and E635 alloys

International Nuclear Information System (INIS)

Smirnov, A.; Markov, D.; Smirnov, V.; Polenok, V.; Ivashchenko, A.; Strozhuk, A.

2006-01-01

The paper presents the main examination results on the condition of fuel rods claddings, guide tubes and spacer grids of the WWER FA made of E110 and E635 alloys operated under standard operating conditions. The paper is based on the data obtained during the examination of 28 WWER-1000 FA and 12 WWER-400 FA. E110 alloy is shown to be suitable material for the WWER fuel rod claddings under the normal operating conditions. E635 alloy is attractive to manufacturing of the skeleton components. The currently used combination (E110 as a material of fuel rods claddings and E635 - as a material of the skeleton components) is the optimal solution for the WWER fuel assembly because the advantages of the both alloys are used. (authors)

Fuel irradiation experience at Halden

International Nuclear Information System (INIS)

Vitanza, Carlo

1996-01-01

The OECD Halden Reactor Project is an international organisation devoted to improved safety and reliability of nuclear power station through an user-oriented experimental programme. A significant part of this programme consists of studies addressing fuel performance issues in a range of conditions realised in specialised irradiation. The key element of the irradiation carried out in the Halden reactor is the ability to monitor fuel performance parameters by means of in-pile instrumentation. The paper reviews some of the irradiation rigs and the related instrumentation and provides examples of experimental results on selected fuel performance items. In particular, current irradiation conducted on high/very high burn-up fuels are reviewed in some detail

The investigation of fast reactor fuel pin start up behaviour in the irradiation experiment DUELL II

International Nuclear Information System (INIS)

Freund, D.; Geithoff, D.

1988-04-01

The irradiation experiments DUELL-II within the SNR-300 operational Transient Experimental Program deal with the investigation of fresh mixed oxide fuel behaviour at start-up. The irradiation has been carried out in the HFR Petten in four so-called DUELL capsules with two fuel pin samples each. The fuel pins with a total length of 453 mm contained a fuel column of 150 mm length, consisting of high dense (U,Pu)O 2-x fuel with an initial porosity of 4%, a Pu-content of 20.9%, and an O/Me ratio of 1.96. The fuel pellet diameter was 6.37 mm, the outer diameter of the SS cladding, material No. 1.4970, was 7.6 mm. The irradiation included four phases, consisting of preconditioning at 85% nominal power (corresponds to 550 W/cm), a following increase to full power, and two following full power periods of 1 and 10 days, respectively. Post irradiation examination showed incomplete fuel restructuring in the first capsules with central void diameters of 800 μm in the hot plane, complete restructuring in the last capsule, leading to central voids of approximately 1 mm diameter. The residual gaps between fuel and clad varied between 25 and 44 μm. The clad inner surface did not show any corrosion attack. The analysis of fuel restructuring has been carried out with the computer code SATURN-S showing good agreement with the PIE results. The analysis led to a series of model improvements, especially for crack volume and relocation modelling. (orig./GL) [de

Estimations on uranium silicide fuel prototypes for their irradiation and postirradiation

International Nuclear Information System (INIS)

Sbaffoni, Maria M.

2000-01-01

The 'Silicide' project includes the qualification of this type of research reactor fuel to be used i.e. in the Argentine RA-3 and to confirm CNEA's role as an international supplier. The present paper shows complementary basic information for P-04 prototype post-irradiation, which is already under way, and some parameter values related to the new P-06 prototype to be taken into account for planning its irradiation and post-irradiation. The reliability of these values has been evaluated through comparison with experimental results. The reported results contribute, also, to a parallel study on the nuclear data libraries used in calculations for this type of reactor. (author)

Evaluation of the characteristics of high burnup and high plutonium content mixed oxide (MOX) fuel

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-08-15

Two kinds of MOX fuel irradiation tests, i.e., MOX irradiation test up to high burnup and MOX having high plutonium content irradiation test, have been performed from JFY 2007 for five years in order to establish technical data concerning MOX fuel behavior during irradiation, which shall be needed in safety regulation of MOX fuel with high reliability. The high burnup MOX irradiation test consists of irradiation extension and post irradiation examination (PIE). The activities done in JFY 2011 are destructive post irradiation examination (D-PIE) such as EPMA and SIMS at CEA (Commissariat a l'Enegie Atomique) facility. Cadarache and PIE data analysis. In the frame of irradiation test of high plutonium content MOX fuel programme, MOX fuel rods with about 14wt % Pu content are being irradiated at BR-2 reactor and corresponding PIE is also being done at PIE facility (SCK/CEN: Studiecentrum voor Kernenergie/Centre d'Etude l'Energie Nucleaire) in Belgium. The activities done in JFY 2011 are non-destructive post irradiation examination (ND-PIE) and D-PIE and PIE data analysis. In this report the results of EPMA and SIMS with high burnup irradiation test and the result of gamma spectrometry measurement which can give FP gas release rate are reported. (author)

Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5.5 at. % burnup

International Nuclear Information System (INIS)

Strain, R.V.; Johnson, C.E.

1978-05-01

Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760 0 C. The maximum diametral change that occurred during irradiation was 0.2% ΔD/D 0 . The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred

Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5. 5 at. % burnup

Energy Technology Data Exchange (ETDEWEB)

Strain, R V; Johnson, C E

1978-05-01

Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760/sup 0/C. The maximum diametral change that occurred during irradiation was 0.2% ..delta..D/D/sub 0/. The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred.

AGC-2 Specimen Post Irradiation Data Package Report

Energy Technology Data Exchange (ETDEWEB)

Windes, William Enoch [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cottle, David L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-08-01

This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One-half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time-of-flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between

Progress in qualifying low-enriched U-Mo dispersion fuels

International Nuclear Information System (INIS)

Snelgrove, J.L.; Hofman, G.L.; Hayes, S.L.; Meyer, M.K.

2001-01-01

The U.S. Reduced Enrichment for Research and Test Reactors program is working to qualify dispersions of U-Mo alloys in aluminum with fuel-meat densities of 8 to 9 gU cm -3 . Post irradiation examinations of the small fuel plates irradiated in the Advanced Test Reactor during the high-temperature RERTR-3 tests are virtually complete, and analysis of the large quantity of data obtained is underway. We have observed that the swelling of the fuel plates is stable and modest and that the swelling is dominated by the temperature-dependent interaction of the U-Mo fuel and the aluminum matrix. In order to extract detailed information about the behavior of these fuels from the data, a complex fuel-plate thermal model is being developed to account for the effects of the changing fission rate and thermal conductivity of the fuel meat during irradiation. This paper summarizes the empirical results of the post irradiation examinations and the preliminary results of the model development. In addition, the schedule for irradiation of full-sized elements in the HFR-Petten is briefly discussed. (author)

Postirradiation examination results for the Irradiation Effects Test 2

International Nuclear Information System (INIS)

Ploger, S.A.; Kerwin, D.K.; Croucher, D.W.

1978-01-01

This report presents the postirradiation examination results of Test IE-2 in the Irradiation Effects Test Series conducted under the Thermal Fuels Behavior Program. The objectives of this test were to evaluate the influence of previous cladding irradiation and fuel-cladding diametral gap on fuel rod behavior during a power ramp and during film boiling operation. Test IE-2, conducted in the Power Burst Facility at the Idaho National Engineering Laboratory, employed two 0.97-m-long pressurized water reactor type fuel rods fabricated from previously irradiated zircaloy-4 cladding and two similar rods fabricated from unirradiated cladding. The four rods were subjected to a preconditioning period, followed by a power ramp to an average peak rod power of 68 kW/m and steady state operation for one hour at an individual rod coolant mass flux of 4880 kg/s . m 2 . After a flow reduction to 2550 kg/s . m 2 , film boiling occurred on three rods. An additional flow reduction to 2245 kg/s . m 2 produced film boiling on the remaining fuel rod. Maximum time in film boiling was 90 s. None of the four fuel rods failed during the test. Damage caused by film boiling, as characterized by oxidation, oxide spalling, and collapse at fuel pellet interfaces, was found on all four rods. Film boiling regions on these rods showed evidence of fuel melting, fuel centerline void formation, and internal cladding oxidation resulting from fuel-cladding reaction. Effects of fuel-cladding diametral gap and cladding irradiation are summarized. Measured temperatures and metallographically estimated temperatures are compared at several axial fuel rod locations

Postirradiation examination results for the Irradiation Effects Test IE-5

International Nuclear Information System (INIS)

Cook, T.F.; Ploger, S.A.; Hobbins, R.R.

1978-03-01

The results are presented of the postirradiation examination of four pressurized water reactor type fuel rods which were tested in-pile under a fast power ramp and film boiling operation during Irradiation Effects (IE) Test 5. The major objectives of this test were to evaluate the effects of simulated fission products on fuel rod behavior during a fast power ramp, to determine the effects of high initial internal pressure on a fuel rod during film boiling, and to assess fuel rod property changes that occur during film boiling in a fuel rod with previously irradiated cladding. The overall condition of the rods and changes that occurred in fuel and cladding as a result of the power ramp and film boiling operation, as determined from the postirradiation examination, are reported and analyzed. Effects of the simulated fission products on fuel rod behavior during a power ramp are discussed. The effect of high internal pressure on rod behavior during film boiling is evaluated. Cladding temperatures are estimated at various axial and circumferential locations. Cladding embrittlement by oxidation is also assessed

Hot fuel examination facility element spacer wire-wrap machine

International Nuclear Information System (INIS)

Tobias, D.A.; Sherman, E.K.

1989-01-01

Nondestructive examinations of irradiated experimental fuel elements conducted in the Argonne National Laboratory Hot Fuel Examination Facility/North (HFEF/N) at the Idaho National Engineering Laboratory include laser and contact profilometry (element diameter measurements), electrical eddy-current testing for cladding and thermal bond defects, bow and length measurements, neutron radiography, gamma scanning, remote visual exam, and photography. Profilometry was previously restricted to spiral profilometry of the element to prevent interference with the element spacer wire wrapped in a helix about the Experimental Breeder Reactor II (EBR-II)-type fuel element from end to end. By removing the spacer wire prior to conducting profilometry examination, axial profilometry techniques may be used, which are considerably faster than spiral techniques and often result in data acquisition more important to experiment sponsors. Because the element must often be reinserted into the nuclear reactor (EBR-II) for additional irradiation, however, the spacer wire must be reinstalled on the highly irradiated fuel element by remote means after profilometry of the wireless elements. The element spacer wire-wrap machine developed at HFEF is capable of helically wrapping fuel elements with diameters up to 1.68 cm (0.660 in.) and 2.44-m (96-in.) lengths. The machine can accommodate almost any desired wire pitch length by simply inserting a new wrapper gear module

Design considerations and operating experience with wet storage of Ontario Hydro's irradiated fuel

International Nuclear Information System (INIS)

Frost, C.R.; Naqvi, S.J.; McEachran, R.A.

1987-01-01

The characteristics of Ontario Hydro's fuel and at-reactor irradiated fuel storage water pools (or irradiated fuel bays) are described. There are two types of bay known respectively as primary bays and auxiliary bays, used for at-reactor irradiated fuel storage. Irradiated fuel is discharged remotely from Ontario Hydro's reactors to the primary bays for initial storage and cooling. The auxiliary bays are used to receive and store fuel after its initial cooling in the primary bay, and provide additional storage capacity as needed. The major considerations in irradiated fuel bay design, including site-specific requirements, reliability and quality assurance, are discussed. The monitoring of critical fuel bay components, such as bay liners, the development of high storage density fuel containers, and the use of several irradiated fuel bays at each reactor site have all contributed to the safe handling of the large quantities of irradiated fuel over a period of about 25 years. Routine operation of the irradiated fuel bays and some unusual bay operational events are described. For safety considerations, the irradiated fuel in storage must retain its integrity. Also, as fuel storage is an interim process, likely for 50 years or more, the irradiated fuel should be retrievable for downstream fuel management phases such as reprocessing or disposal. A long-term experimental program is being used to monitor the integrity of irradiated fuel in long-term wet storage. The well characterized fuel, some of which has been in wet storage since 1962 is periodically examined for possible deterioration. The evidence from this program indicates that there will be no significant change in irradiated fuel integrity (and retrievability) over a 50 year wet storage period

UO2 fuel behaviour at rod burn-ups up to 105 MWd/kgHM. A review of 10 years of high burn-up examinations commissioned by AREVA NP

International Nuclear Information System (INIS)

Goll, W.; Hoffmann, P.B.; Hellwig, C.; Sauser, W.; Spino, J.; Walker, C.T.

2007-01-01

Irradiation experience gained on fuel rods with burn-ups greater than 60 MWd/kgHM irradiated in the Nuclear Power Plant Goesgen, Switzerland, is described. Emphasis is placed on the fuel behaviour, which has been analysed by hot cell examinations at the Institute for Transuranium Elements and the Paul-Scherrer-Institute. Above 60 MWd/kgHM, the so-called high burn-up structure (HBS) forms and the fission gas release increases with burn-up and rod power. Examinations performed in the outer region of the fuel revealed that most if not all of the fission gas created was retained in the HBS, even at 25% porosity. Furthermore, the HBS has a relatively low swelling rate, greatly increased plasticity, and its thermal conductivity is higher than expected from the porosity. The post-irradiation examinations showed that the HBS has no detrimental effects on the performance of stationary irradiated PWR fuel irradiated to the high burn-ups that can be achieved with 5 wt% U-235 enrichment. On the contrary, the HBS results in fuel performance that is generally better than it would have been if the HBS had not formed. (orig.)

A Study on Structural Strength of Irradiated Spacer Grid for PWR Fuel

Energy Technology Data Exchange (ETDEWEB)

Jin, Y. G.; Baek, S. J.; Kim, D. S.; Yoo, B. O.; Ahn, S. B.; Chun, Y. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, J. I.; Kim, Y. H.; Lee, J. J. [KEPCO NF, Daejeon (Korea, Republic of)

2014-10-15

A fuel assembly consists of an array of fuel rods, spacer grids, guide thimbles, instrumentation tubes, and top and bottom nozzles. In PWR (Pressurized light Water Reactor) fuel assemblies, the spacer grids support the fuel rods by the friction forces between the fuel rods and springs/dimples. Under irradiation, the spacer grids supporting the fuel rods absorb vibration impacts due to the reactor coolant flow, and also bear static and dynamic loads during operation inside the nuclear reactor and transportation for spent fuel storage. Thus, it is important to understand the characteristics of deformation behavior and the change in structural strength of an irradiated spacer grid.. In the present study, the static compression test of a spacer grid was conducted to investigate the structural strength of the irradiated spacer grid in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. To evaluate the structural strength of an irradiated spacer grid, hot cell tests were carried out at IMEF of KAERI. The fuel assembly was dismantled and the irradiated spacer grid was obtained for the compression test. The apparatus for measuring the compression strength of the irradiated spacer grid was developed and installed successfully in the hot cell.

Post Irradiation Examination Results of the NT-02 Graphite Fins NUMI Target

Energy Technology Data Exchange (ETDEWEB)

Ammigan, K. [Fermilab; Hurh, P. [Fermilab; Sidorov, V. [Fermilab; Zwaska, R. [Fermilab; Asner, D. M. [PNL, Richland; Casella, Casella,A.M [PNL, Richland; Edwards, D. J. [PNL, Richland; Schemer-Kohrn, A. L. [PNL, Richland; Senor, D. J. [PNL, Richland

2017-02-10

The NT-02 neutrino target in the NuMI beamline at Fermilab is a 95 cm long target made up of segmented graphite fins. It is the longest running NuMI target, which operated with a 120 GeV proton beam with maximum power of 340 kW, and saw an integrated total proton on target of 6.1 1020. Over the last half of its life, gradual degradation of neutrino yield was observed until the target was replaced. The probable causes for the target performance degradation are attributed to radiation damage, possibly including cracking caused by reduction in thermal shock resistance, as well as potential localized oxidation in the heated region of the target. Understanding the long-termstructural response of target materials exposed to proton irradiation is critical as future proton accelerator sources are becoming increasingly more powerful. As a result, an autopsy of the target was carried out to facilitate post-irradiation examination of selected graphite fins. Advanced microstructural imaging and surface elemental analysis techniques were used to characterize the condition of the fins in an effort to identify degradation mechanisms, and the relevant findings are presented in this paper.

INFLUENCE OF FUEL-MATRIX INTERACTION ON THE BREAKAWAY SWELLING OF U-MO DISPERSION FUEL IN AL

OpenAIRE

HO JIN RYU; YEON SOO KIM

2014-01-01

In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model prediction...

Cerenkov methodology for monitoring irradiated reactor fuel

International Nuclear Information System (INIS)

Nicholson, N.; Dowdy, E.J.

1984-01-01

Attribute measurement methods for confirming declared irradiated fuel inventories at nuclear installations under safeguards surveillance are of significant interest to inspectors. High-gain measurements of the intensity of the Cerenkov glow from exposed assemblies in water-filled storage ponds are promising for this purpose because the measured intensities depend on cooling times and burnup. We have developed a Cerenkov Measuring Device, a hand-held instrument that examines irradiated fuel assemblies in water-filled storage ponds and measures the intensity of the associated Cerenkov glow. In addition, we have developed a method for making such high-gain measurements in the presence of intense ambient light

Post irradiation examination and experience

International Nuclear Information System (INIS)

1985-11-01

The present meeting was scheduled by the International Atomic Energy Agency upon proposal from the members of the International Working Group on Water Reactor Fuel Performance and Technology. At the invitation of the Government of Japan, Japan Atomic Energy Research Institute and Nuclear Safety Research Association (of Japan), organized the meeting in Tokyo. 37 participants representing 13 countries and one international organization attended the meeting. 27 papers were presented in three sessions, namely: general fuel testing programme (3 papers), fuel performance study (10 papers), in-core, on-site and hot cell technique (8 papers). A separate abstract was prepared for each of these papers. Three syndicate meetings allowed participants to discuss the papers and to draw up conclusions and recommendations

Segmented fuel irradiation program: investigation on advanced materials

International Nuclear Information System (INIS)

Uchida, H.; Goto, K.; Sabate, R.; Abeta, S.; Baba, T.; Matias, E. de; Alonso, J.

1999-01-01

The Segmented Fuel Irradiation Program, started in 1991, is a collaboration between the Japanese organisations Nuclear Power Engineering Corporation (NUPEC), the Kansai Electric Power Co., Inc. (KEPCO) representing other Japanese utilities, and Mitsubishi Heavy Industries, Ltd. (MHI); and the Spanish Organisations Empresa Nacional de Electricidad, S.A. (ENDESA) representing A.N. Vandellos 2, and Empresa Nacional Uranio, S.A. (ENUSA); with the collaboration of Westinghouse. The objective of the Program is to make substantial contribution to the development of advanced cladding and fuel materials for better performance at high burn-up and under operational power transients. For this Program, segmented fuel rods were selected as the most appropriate vehicle to accomplish the aforementioned objective. Thus, a large number of fuel and cladding combinations are provided while minimising the total amount of new material, at the same time, facilitating an eventual irradiation extension in a test reactor. The Program consists of three major phases: phase I: design, licensing, fabrication and characterisation of the assemblies carrying the segmented rods (1991 - 1994); phase II: base irradiation of the assemblies at Vandellos 2 NPP, and on-site examination at the end of four cycles (1994-1999). Phase III: ramp testing at the Studsvik facilities and hot cell PIE (1996-2001). The main fuel design features whose effects on fuel behaviour are being analysed are: alloy composition (MDA and ZIRLO vs. Zircaloy-4); tubing texture; pellet grain size. The Program is progressing satisfactorily as planned. The base irradiation is completed in the first quarter of 1999, and so far, tests and inspections already carried out are providing useful information on the behaviour of the new materials. Also, the Program is delivering a well characterized fuel material, irradiated in a commercial reactor, which can be further used in other fuel behaviour experiments. The paper presents the main

Irradiation test plan of instrumented capsule(05F-01K) for nuclear fuel irradiation in Hanaro (Revision 1)

Energy Technology Data Exchange (ETDEWEB)

Sohn, Jae Min; Kim, B. G.; Choi, M. H. (and others)

2006-09-15

An instrumented capsule was developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel pellet elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.84 full power days at 24 MW). In the year of 2004, 3 test fuel rods and the 03F-05K instrumented fuel capsule were designed and fabricated to measure fuel centerline temperature, internal pressure of fuel rod, and fuel axial deformation during irradiation test. Now, this capsule was successfully irradiated in the test hole OR5 of HANARO reactor from April 27, 2004 to October 1, 2004 (59.5 full power days at 24-30 MW). The capsule and fuel rods have been be dismantled and fuel rods have been examined at the hot cell of IMEF. The instrumented fuel capsule (05F-01K) was designed and manufactured for a design verification test of the dual instrumented fuel rods. The irradiation test of the 05F-01K instrumented fuel capsule will be carried out at the OR5 vertical experimental hole of HANARO.

A data recording and processing system for examination of irradiated fuel rods

International Nuclear Information System (INIS)

Olshausen, K.D.; Christiansen, J.F.; Larsen, J.K.; Stevens, T.

1978-03-01

Over the past eight years a system has been developed at Institutt for atomenergi, Kjeller, Norway, which helps to collect and analyze data obtained during certain post-irradiation examination techniques employed at IFA's Hot Laboratory. At present computer codes are in use for treating results from: profilometry, gamma scanning, eddy current testing and tensile testing. The system is easily adaptable to any othermeasuring data with the structure: -axial position, time interval etc.; signal A (DC-voltage, counts); signal B (DC-voltage, counts). Anextention of the system for transverse gamma scanning and gap measurement (mechanical compression method) is planned for the future. The various programmes in the system are described and examples of the outputs are given in a series of appendices. (JIW)

A Study on Cell Size of Irradiated Spacer Grid for PWR Fuel

International Nuclear Information System (INIS)

Jin, Y. G.; Kim, G. S.; Ryu, W. S. and others

2014-01-01

The spacer grids supporting the fuel rods absorb vibration impacts due to the reactor coolant flow, and grid spring force decreases under irradiation. This reduction of contact force might cause grid-to-rod fretting wear. The fretting failure of the fuel rod is one of the recent significant issues in the nuclear industry from an economical as well as a safety concern. Thus, it is important to understand the characteristics of cell spring behavior and the change in size of grid cells for an irradiated spacer grid. In the present study, the dimensional measurement of a spacer grid was conducted to investigate the cell size of an irradiated spacer grid in a hot cell at IMEF (Irradiated Materials Examination Facility) of KAERI. To evaluate the fretting wear performance of an irradiated spacer grid, hot cell tests were carried out at IMEF of KAERI. Hot cell examinations include dimensional measurements for the irradiated spacer grid. The change of cell sizes was dependent on the direction of the spacer grids, leading to significant gap variations. It was found that the change in size of the cell springs due to irradiation-induced stress relaxation and creep during the fuel residency in the reactor core affect the contact behavior between the fuel rod and the cell spring

Safeguards approach for irradiated fuel

International Nuclear Information System (INIS)

Harms, N.L.; Roberts, F.P.

1987-03-01

IAEA verification of irradiated fuel has become more complicated because of the introduction of variations in what was once presumed to be a straightforward flow of fuel from reactors to reprocessing plants, with subsequent dissolution. These variations include fuel element disassembly and reassembly, rod consolidation, double-tiering of fuel assemblies in reactor pools, long term wet and dry storage, and use of fuel element containers. This paper reviews future patterns for the transfer and storage of irradiated LWR fuel and discusses appropriate safeguards approaches for at-reactor storage, reprocessing plant headend, independent wet storage, and independent dry storage facilities

Influence of fuel-matrix interaction on the breakaway swelling of U-Mo dispersion fuel in Al

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Yeon Soo [Nuclear Engineering Division, Argonne National Laboratory, Arogonne (United States)

2014-04-15

In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model predictions, advantageous fuel design parameters are recommended to prevent breakaway swelling.

Plutonium fuel program

International Nuclear Information System (INIS)

1979-09-01

A review is presented of the development of the (UPu)C sphere-pac fuel project during 1978. In particular, the problems encountered in obtaining good fuel quality in the fabrication process and their solution is discussed. The development of a fabrication pilot plant is considered, and the post-irradiation examination of fuel pins is presented. (Auth.)

UMo nuclear fuels behaviour under heavy ion irradiation: a μ-XAS study

International Nuclear Information System (INIS)

Palancher, H.; Martin, P.; Dubois, S.; Valot, C.; Sabathier, C.; Palancher, H.; Nassif, V.; Proux, O.; Hazemann, J.L.; Wieschalla, N.; Petry, W.; Jarousse, C.

2007-01-01

Full text of publication follows. A worldwide program encourages the use of low enriched uranium (LEU, 235 U 235 U concentration up to 93 wt. %). Due to the decrease in 235 U enrichment for the conversion to LEU, the total density of uranium atoms in the fuel must be increased accordingly. To preserve the neutron flux, metallic uranium alloys could be the best fuel material. The fuel, which consists of UMo alloy spherical particles surrounded by an Al matrix (cf. Figure 1-A), is rolled between two aluminium claddings. Post-irradiation examinations of U-7 wt%Mo demonstrated its strong potentialities as fuel but they also pointed out its interaction with aluminium (cf. Figure 1-B). In certain cases this interaction can cause a break-away swelling of the plate. The aim of this project is the understanding of: - the phenomena driving the growth of the interaction layer. - the influence on interaction layer composition of limited adjunction of elements (silicon...) to the Al matrix. To overcome the difficulties inherent to the in-pile irradiated samples, an out-of-pile methodology (collaboration between CEA, FRM II and CERCA) has been developed based on heavy ion irradiation. This methodology enables to simulate the fission fragment damages using a 80 MeV iodine beam at the Maier Leibnitz laboratory (Garching, Germany). After irradiation, samples are characterised at micrometer scale by microscopy (SEM coupled with EDX) and X-Ray techniques (XRD and XAS). The irradiation (final dose: 2 x 10 17 at/cm 2 ) of undoped U-7 wt%Mo fuel plates leads to the formation of an interaction layer surrounding each fuel particles (cf. Figure 1-C). μ-XRD analysis performed at the ESRF (ID18f) showed only the presence of UAl 3 phase in the interaction layer. Same results have been obtained on in-pile irradiated fuel by Sears et al using neutron diffraction confirming the interest of the developed methodology. However the behaviour of the Mo atoms in the interaction layer could not be

Determination of burn-up of irradiated nuclear fuels using mass spectrometry

International Nuclear Information System (INIS)

Jagadish Kumar, S.; Telmore, V.M.; Shah, R.V.; Sasi Bhushan, K.; Paul, Sumana; Kumar, Pranaw; Rao, Radhika M.; Jaison, P.G.

2017-01-01

Burn-up defined as the atom percent fission, is a vital parameter used for assessing the performance of nuclear fuel during its irradiation in the reactor. Accurate data on the actinide isotopes are also essential for the reliable accountability of nuclear materials and for nuclear safeguards. Both destructive and non-destructive methods are employed in the post-irradiation analysis for the burn-up measurements. Though non-destructive methods are preferred from the point view of remote handling of irradiated fuels with high radioactivity, they do not provide the high accuracy as achieved by the chemical analysis methods. Thus destructive radiochemical and chemical analyses are still the established reference methods for accurate and reliable burn-up determination of irradiated nuclear fuels. In the destructive method, burn-up of irradiated nuclear fuel is determined by correlating the amount of a fission product formed during irradiation with that of heavy elements. Thus the destructive experimental determination of burn-up involves the dissolution of irradiated fuel samples followed by the separation and determination of heavy elements and fission product(s) to be used as burn-up monitor(s). Another approach for the experimental determination of burn-up is based on the changes in the abundances of the heavy element isotopes. A widely accepted method for burn-up determination is based on stable "1"4"8Nd and "1"3"9La as burn-up monitors. Several properties such as non-volatility, nearly same yields for thermal fissions of "2"3"5U and "2"3"9Pu etc justifies the selection of "1"4"8Nd as a burn-up monitor

Post-Irradiation Examination of Array Targets - Part I

Energy Technology Data Exchange (ETDEWEB)

Icenhour, A.S.

2004-01-23

During FY 2001, two arrays, each containing seven neptunium-loaded targets, were irradiated at the Advanced Test Reactor in Idaho to examine the influence of multi-target self-shielding on {sup 236}Pu content and to evaluate fission product release data. One array consisted of seven targets that contained 10 vol% NpO{sub 2} pellets, while the other array consisted of seven targets that contained 20 vol % NpO{sub 2} pellets. The arrays were located in the same irradiation facility but were axially separated to minimize the influence of one array on the other. Each target also contained a dosimeter package, which consisted of a small NpO{sub 2} wire that was inside a vanadium container. After completion of irradiation and shipment back to the Oak Ridge National Laboratory, nine of the targets (four from the 10 vol% array and five from the 20 vol% array) were punctured for pressure measurement and measurement of {sup 85}Kr. These nine targets and the associated dosimeters were then chemically processed to measure the residual neptunium, total plutonium production, {sup 238}Pu production, and {sup 236}Pu concentration at discharge. The amount and isotopic composition of fission products were also measured. This report provides the results of the processing and analysis of the nine targets.

Postirradiation examination of high-density uranium alloy dispersion fuels

International Nuclear Information System (INIS)

Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Strain, R.V.

1998-01-01

Two irradiation test vehicles, designated RERTR-2, were inserted into the Advanced Test reactor in Idaho in August 1997. These tests were designed to obtain irradiation performance information on a variety of potential new, high-density uranium alloy dispersion fuels, including U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru and U-10Mo-0.05Sn: the intermetallic compounds U 2 Mo and U-10Mo-0.-5Sn; the intermetallic compounds U 2 Mo and U 3 Si 2 were also included in the fuel test matrix. These fuels are included in the experiments as microplates (76 mm x 22 mm x 1.3mm outer dimensions) with a nominal fuel volume loading of 25% and irradiated at relatively low temperature (∼100 deg C). RERTR-1 and RERTR-2 were discharged from the reactor in November 1997 and July 1998, respectively at calculated peak fuel burnups of 45 and 71 at %-U 235 Both experiments are currently under examination at the Alpha Gamma Hot Cell Facility at Argonne National Laboratory in Chicago. This paper presents the postirradiation examination results available to date from these experiments. (author)

Final report on development and operation of instrumented irradiation capsules for creep experiments on nuclear fuels at FR2

International Nuclear Information System (INIS)

Haefner, H.E.; Philipp, K.; Blumhofer, M.

1980-02-01

The capsule test rig No. 154 removed from FR2 in April 1979 was the last irradiation rig in a long series of creep experiments. The target of the irradiation tests, started exactly ten years ago, was to investigate the creep behaviour of various ceramic nuclear fuels under different in-pile irradiation conditions. An irradiation test rig had been developed for this purpose which allowed the continuous measurement of changes in length of fuel specimens. A total of 28 capsule test rigs each containing two packages of creep specimens have been irradiated in FR2 during this decade. They included 23 specimen stacks of UO 2 , 16 specimen stacks of UO 2 -PuO 2 , 4 specimen stacks of UN, 10 specimen stacks of (U,Pu) C, and 13 reference specimens of molybdenum. Besides the description of the test facility, the report provides above all a survey of the operation data applicable to the specimens and of the operating experience gathered as well as of the findings obtained in post-irradiation examinations. (orig.) [de

Postirradiation examination of Peach Bottom HTGR Driver Fuel Element E06-01

International Nuclear Information System (INIS)

Dyer, F.F.; Wichner, R.P.; Martin, W.J.; Fairchild, L.L.; Kedl, R.J.; de Nordwall, H.J.

1976-04-01

The report presented describes the postirradiation examinations of driver fuel element E06-01, which had been irradiated an equivalent of 384 full-power days in Peach Bottom, Unit 1. The fuel element is described in detail and its temperature and irradiation service history briefly outlined. Results presented include: (1) visual observations; (2) critical dimensions of fuel compacts, sleeve, and spine; (3) axial distributions of gamma-emitting nuclides plus 3 H and 90 Sr; (4) radial distributions of these nuclides in the sleeve and spine at three axial locations in the fueled regions and three locations in the upper reflector; (5) metallographic examination of samples of fuel compact material; and (6) burnup determinations via radiochemical analyses at two compact locations

Dry Process Fuel Performance Evaluation

Energy Technology Data Exchange (ETDEWEB)

Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

2005-04-15

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

Dry Process Fuel Performance Evaluation

International Nuclear Information System (INIS)

Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

2005-04-01

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

Small-scale irradiated fuel electrorefining

International Nuclear Information System (INIS)

Benedict, R.W.; Krsul, J.R.; Mariani, R.D.; Park, K.; Teske, G.M.

1993-01-01

In support of the metallic fuel cycle development for the Integral Fast Reactor (IFR), a small scale electrorefiner was built and operated in the Hot Fuel Examination Facility (HFEF) at Argonne National Laboratory-West. The initial purpose of this apparatus was to test the single segment dissolution of irradiated metallic fuel via either direct dissolution in cadmium or anodic dissolution. These tests showed that 99.95% of the uranium and 99.99% of the plutonium was dissolved and separated from the fuel cladding material. The fate of various fission products was also measured. After the dissolution experiments, the apparatus was upgraded to stady fission product behavior during uranium electrotransport. Preliminary decontamination factors were estimated for different fission products under different processing conditions. Later modifications have added the following capabilities: Dissolution of multiple fuel segments simultaneously, electrotransport to a solid cathode or liquid cathode and actinide recovery with a chemical reduction crucible. These capabilities have been tested with unirradiated uranium-zirconium fuel and will support the Fuel Cycle Demonstration program

Irradiation test plan of the simulated DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Bae, Ki Kwang; Yang, M. S.; Kim, B. K. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-11-01

Simulated DUPIC fuel had been irradiated from Aug. 4, 1999 to Oct. 4 1999, in order to produce the data of its in-core behavior, to verify the design of DUPIC non-instrumented capsule developed, and to ensure the irradiation requirements of DUPIC fuel at HANARO. The welding process was certified for manufacturing the mini-element, and simulated DUPIC fuel rods were manufactured with simulated DUPIC pellets through examination and test. The non-instrumented capsule for a irradiation test of DUPIC fuel has been designed and manufactured referring to the design specification of the HANARO fuel. This is to be the design basis of the instrumented capsule under consideration. The verification experiment, whether the capsule loaded in the OR4 hole meet the HANARO requirements under the normal operation condition, as well as the structural analysis was carried out. The items for this experiment were the pressure drop test, vibration test, integrity test, et. al. It was noted that each experimental result meet the HANARO operational requirements. For the safety analysis of the DUPIC non-instrumented capsule loaded in the HANARO core, the nuclear/mechanical compatibility, thermodynamic compatibility, integrity analysis of the irradiation samples according to the reactor condition as well as the safety analysis of the HANARO were performed. Besides, the core reactivity effects were discussed during the irradiation test of the DUPIC capsule. The average power of each fuel rod in the DUPIC capsule was calculated, and maximal linear power reflecting the axial peaking power factor from the MCNP results was evaluated. From these calculation results, the HANARO core safety was evaluated. At the end of this report, similar overseas cases were introduced. 9 refs., 16 figs., 10 tabs. (Author)

New high density MTR fuel. The CEA-CERCA-COGEMA development program

International Nuclear Information System (INIS)

Languille, A.; Durand, J.P.; Gay, A.

1999-01-01

The development of a new generation of LEU, high in density and with reprocessing capacities MTR fuel, is a key issue to provide reactor operators with a smooth operation which is necessary for a long term development of Nuclear Energy. In the RRFM'98 meeting, a joint contribution of CEA, CERCA and COGEMA presented a technical classification of the potential candidates uranium alloys. In this paper this MTR working group presents the development program of a new high density fuel. This program is composed of three main steps: Basic Data analysis and collection, Plate Tests (Irradiation and Post Irradiation Examinations) and Lead Test Assemblies (Irradiation and Post Irradiation Examinations). The goal to be reached is to make this new fuel available before the end of the present US return policy. (author)

Irradiated fuel performance evaluation technology development

International Nuclear Information System (INIS)

Koo, Yang Hyun; Bang, J. G.; Kim, D. H.

2012-01-01

Alpha version performance code for dual-cooled annular fuel under steady state operation, so called 'DUOS', has been developed applying performance models and proposed methodology. Furthermore, nonlinear finite element module which could be integrated into transient/accident fuel performance code was also developed and evaluated using commercial FE code. The first/second irradiation and PIE test of annular pellet for dual-cooled annular fuel in the world have been completed. In-pile irradiation test DB of annular pellet up to burnup of 10,000 MWd/MTU through the 1st test was established and cracking behavior of annular pellet and swelling rate at low temperature were studied. To do irradiation test of dual-cooled annular fuel under PWR's simulating steady-state conditions, irradiation test rig/rod design/manufacture of mock-up/performance test have been completed through international collaboration program with Halden reactor project. The irradiation test of large grain pellets has been continued from 2002 to 2011 and completed successfully. Burnup of 70,000 MWd/MTU which is the highest burnup among irradiation test pellets in domestic was achieved

Development and verifications of fast reactor fuel design code ''Ceptar''

International Nuclear Information System (INIS)

Ozawa, T.; Nakazawa, H.; Abe, T.

2001-01-01

The annular fuel is very beneficial for fast reactors, because it is available for both high power and high burn-up. Concerning the irradiation behavior of the annular fuel, most of annular pellets irradiated up to high burn-up showed shrinkage of the central hole due to deformation and restructuring of the pellets. It is needed to predict precisely the shrinkage of the central hole during irradiation, because it has a great influence on power-to-melt. In this paper, outline of CEPTAR code (Calculation code to Evaluate fuel pin stability for annular fuel design) developed to meet this need is presented. In this code, the radial profile of fuel density can be computed by using the void migration model, and law of conservation of mass defines the inner diameter. For the mechanical analysis, the fuel and cladding deformation caused by the thermal expansion, swelling and creep is computed by the stress-strain analysis using the approximation of plane-strain. In addition, CEPTAR can also take into account the effect of Joint-Oxide-Gain (JOG) which is observed in fuel-cladding gap of high burn-up fuel. JOG has an effect to decrease the fuel swelling and to improve the gap conductance due to deposition of solid fission product. Based on post-irradiation data on PFR annular fuel, we developed an empirical model for JOG. For code verifications, the thermal and mechanical data obtained from various irradiation tests and post-irradiation examinations were compared with the predictions of this code. In this study, INTA (instrumented test assembly) test in JOYO, PTM (power-to-melt) test in JOYO, EBR-II, FFTF and MTR in Harwell laboratory, and post-irradiation examinations on a number of PFR fuels, were used as verification data. (author)

Remote waste handling at the Hot Fuel Examination Facility

International Nuclear Information System (INIS)

Vaughn, M.E.

1982-01-01

Radioactive solid wastes, some of which are combustible, are generated during disassembly and examination of irradiated fast-reactor fuel and material experiments at the Hot Fuel Examination Facility (HFEF). These wastes are remotely segregated and packaged in doubly contained, high-integrity, clean, retrievable waste packages for shipment to the Radioactive Waste Management Complex (RWMC) at the Idaho National Engineering Laboratory (INEL). This paper describes the equipment and techniques used to perform these operations

Status of LEU fuel development and conversion of NRU

International Nuclear Information System (INIS)

Sears, D.F.; Herbert, L.N.; Vaillancourt, K.D.

1989-11-01

The status of the low-enrichment uranium (LEU) fuel development and NRU conversion program at Chalk River Nuclear Laboratories is reviewed. Construction of a new fuel fabrication facility is essentially completed and installation of LEW fuel manufacturing equipment has begun. The irradiation of 31 prototype Al-61 wt% U 3 Si dispersion fuel rods, approximately one third of a full NRU core, is continuing without incident. Recent post-irradiation examination of spent fuel rods revealed that the prototype LEU fuel achieved the design burnup (80 at%) in excellent condition, confirming that the Al-U 3 Si 2 dispersion fuel to complement out Al-U 3 Si capability. Three full-size NRU rods containing Al-U 3 Si 2 dispersion fuel have been fabricated for a qualification irradiation in NRU. Post-irradiation examinations of mini-elements containing Al-U 3 Si 2 fuel revealed that the U 3 Si 2 behaved similarly to U 3 Si 2 fuel revealed that the U 3 Si 2 particles and the aluminum matrix, and fission gas bubbles up to 10 μm in diameter, could be seen in the particles after 60 at% and 80 at% burnup. The mini-elements contained a variety of silicide particle sizes; however, no significant swelling dependence on particle size distribution was observed

Fuel temperature prediction during high burnup HTGR fuel irradiation test. US-JAERI irradiation test for HTGR fuel

International Nuclear Information System (INIS)

Sawa, Kazuhiro; Fukuda, Kousaku; Acharya, R.

1995-01-01

This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for an irradiation test in a removable beryllium position of the High Flux Isotope Reactor(HFIR) at Oak Ridge National Laboratory. This test is being carried out under Annex 2 of the Arrangement between the U.S. Department of Energy and the Japan Atomic Energy Research Institute on Cooperation in Research and Development regarding High-Temperature Gas-cooled Reactors. The fuel used in the test is an advanced type. The advanced fuel was designed aiming at burnup of about 10%FIMA(% fissions per initial metallic atom) which was higher than that of the first charge fuel for the High Temperature Engineering Test Reactor(HTTR) and was produced in Japan. CACA-2, a heavy isotope and fission product concentration calculational code for experimental irradiation capsules, was used to determine time-dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries(HEATING) code was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body, which contains the fuel compacts, and of the primary pressure vessel were determined such that the requirements of running the fuel compacts at an average temperature less than 1250degC and of not exceeding a maximum fuel temperature of 1350degC were met throughout the four cycles of irradiation. The detail design of the capsule was carried out based on this analysis. (author)

Nuclear fuel

International Nuclear Information System (INIS)

D Hondt, P.

1998-01-01

The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

Dry process fuel performance technology development

International Nuclear Information System (INIS)

Kang, Kweon Ho; Kim, K. W.; Kim, B. K.

2006-06-01

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

Dry process fuel performance technology development

Energy Technology Data Exchange (ETDEWEB)

Kang, Kweon Ho; Kim, K. W.; Kim, B. K. (and others)

2006-06-15

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

Main examination results of WWER-1000 fuel after its irradiation in power reactors

Energy Technology Data Exchange (ETDEWEB)

Bibiliashvili, Yu [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation); Dubrovin, K [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Vasilchenko, I [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation); Yenin, A; Kushmanov, A [AO Novosibirskij Zavod Khimcontsentratov, Novosibirsk (Russian Federation); Smirnov, A; Smirnov, V [Nauchno-Issledovatel` skij Inst. Atomnykh Reaktorov, Dimitrovgrad (Russian Federation)

1994-12-31

WWER-1000 fuel examination has been undertaken to specify the properties of fuel assembly members by defining the parameters of their materials and their interconnection in power reactor operation conditions. Nine fuel assemblies are examined. The examination program includes: visual inspection, measurement of overall dimensions, eddy-current test, gamma-scanning, X-ray and neutron radiography, analysis of gas pressure and composition inside fuel rods, ceramography/metallography, mass spectrometry, microanalysis and electron microscopy of fuel and fuel claddings. The examination results suggest that WWER-1000 fuel spent at steady-state operation conditions up to 50 Mwd/kg U of burnup is in satisfactory condition. The examination of all types of fuel cladding failures indicates that the reason lies in the interaction of cladding with coolant solid impurities. The nodular cladding corrosion of fuel assembly discharged from the South-Ukrainian NPP is caused by the graphite compounds deposited on the fuel rod. Those deposits are a result of the circulating pump damage and had accidental, non-typical character. Some of the rods were found to have a small cladding `fretting` of the spacer grid cell material. The values of the majority of parameters determining the fuel efficiency allow to assume that there is a potential for further extension of fuel burnup and operation length. 1 tab., 11 figs.

Main examination results of WWER-1000 fuel after its irradiation in power reactors

International Nuclear Information System (INIS)

Bibiliashvili, Yu.; Dubrovin, K.; Vasilchenko, I.; Yenin, A.; Kushmanov, A.; Smirnov, A.; Smirnov, V.

1994-01-01

WWER-1000 fuel examination has been undertaken to specify the properties of fuel assembly members by defining the parameters of their materials and their interconnection in power reactor operation conditions. Nine fuel assemblies are examined. The examination program includes: visual inspection, measurement of overall dimensions, eddy-current test, gamma-scanning, X-ray and neutron radiography, analysis of gas pressure and composition inside fuel rods, ceramography/metallography, mass spectrometry, microanalysis and electron microscopy of fuel and fuel claddings. The examination results suggest that WWER-1000 fuel spent at steady-state operation conditions up to 50 Mwd/kg U of burnup is in satisfactory condition. The examination of all types of fuel cladding failures indicates that the reason lies in the interaction of cladding with coolant solid impurities. The nodular cladding corrosion of fuel assembly discharged from the South-Ukrainian NPP is caused by the graphite compounds deposited on the fuel rod. Those deposits are a result of the circulating pump damage and had accidental, non-typical character. Some of the rods were found to have a small cladding 'fretting' of the spacer grid cell material. The values of the majority of parameters determining the fuel efficiency allow to assume that there is a potential for further extension of fuel burnup and operation length. 1 tab., 11 figs

Observation on the irradiation behavior of U-Mo alloy dispersion fuel

International Nuclear Information System (INIS)

Hofman, Gerard L.; Meyer, Mitchell K.; Park, Jong-Man

2000-01-01

Initial results from the postirradiation examination of high-density dispersion fuel test RERTR-3 are discussed. The U-Mo alloy fuels in this test were irradiated to 40% U-235 burnup at temperature ranging from 140 0 C to 240 0 C. Temperature has a significant effect on overall swelling of the test plates. The magnitude of the swelling appears acceptable and no unstable irradiation behavior is evident. (author)

Predictions of the thermomechanical code ''RESTA'' compared with fuel element examinations after irradiation in the BR3 reactor

International Nuclear Information System (INIS)

Petitgrand, S.

1980-01-01

A large number of fuel rods have been irradiated in the small power plant BR3. Many of them have been examined in hot cells after irradiation, giving thus valuable experimental information. On the other hand a thermomechanical code, named RESTA, has been developed by the C.E.A. to describe and predict the behaviour of a fuel pin in a PWR environment and in stationary conditions. The models used in that code derive chiefly from the C.E.A.'s own experience and are briefly reviewed in this paper. The comparison between prediction and experience has been performed for four power history classes: (1) moderate (average linear rating approximately equal to 20 kw m -1 ) and short (approximately equal to 300 days) rating, (2) moderate (approximately equal to 20 kw m -1 ) and long (approximately equal to 600 days) rating, (3) high (25-30 kw m -1 ) and long (approximately equal to 600 days) rating and (4) very high (30-40 kw m -1 ) and long (approximately equal to 600 days) rating. Satisfactory agreement has been found between experimental and calculated results in all cases, concerning fuel structural change, fission gas release, pellet-clad interaction as well as clad permanent strain. (author)

The physics of irradiated nuclear fuel

International Nuclear Information System (INIS)

Robin, M.

1980-01-01

The knowledge of the neutron irradiation effect is essential in dealing with all subjects related to the fuel. Neutron irradiation provokes fission reactions within the fuel and produces new nuclides. The formation chains are described and the importance of each isotope in the fuel cycle is explained with regards to its own characteristics. To solve the system of equations giving the evolution of different nuclides concentrations, the corresponding effective cross-sections and flux received are given by standard codes used for reactor calculations. A good test for calculation methods is the experimental study of irradiated fuel. Many techniques have been developed for this purpose. The last chapter compares fuel evolution in different reactors, in connection with some specific characteristics. (author)

Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti

Energy Technology Data Exchange (ETDEWEB)

Horhoianu, Grigore [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.; Sorescu, Ion [Institute for Nuclear Research, Pitesti (Romania). TRIGA Reactor Loop Facility; Parvan, Marcel [Institute for Nuclear Research, Pitesti (Romania). Hot Cells Lab.

2012-12-15

Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO{sub 2} grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation. (orig.)

Post-irradiation examinations and theoretical analyses of the fuel and fuel pin behavior of innovative systems

International Nuclear Information System (INIS)

Ernst, W.; Freund, D.; Geithoff, D.; Heck, M.; Jacobi, O.; Steiner, H.; Weimar, P.

1995-01-01

For the development of LMR cladding materials an irradiation test was performed in the KNK II reactor between 1986 and 1991. It demonstrated that a temperature gradient in the cladding has no observable effect on swelling behavior. PIEs of transient experiments in the Petten HFR have shown no significant cladding deformations. (orig.)

Recent status of development and irradiation performance for plate type fuel elements with reduced 235U enrichment at NUKEM

International Nuclear Information System (INIS)

Hrovat, M.F.; Hassel, H.W.

1984-01-01

According to the present state of development full size test fuel elements with the maximum uranium densities of 2,2 g U/cm 3 meat for UAlsub(x), 3,2 g U/cm 3 meat for U 3 O 8 and 4,8 g U/cm 3 meat for U 3 Si 2 can be fabricated at NUKEM in production scale. Special chemical procedures for the uranium recovery were developed ensuring an economic fuel fabrication process. The post irradiation examinations (PIE) of 12 UAlsub(x) (U density 2,2 g U/cm 3 meat) and U 3 O 8 (up to 3,1 g U/cm 3 meat) test plates irradiated in the ORR, Oak Ridge research reactor, were terminated. All 12 test plates show unobjectionable irradiation behavior. Extensive irradiation tests on full size fuel elements were performed. All inserted elements show perfect irradiation behavior. The PIE of the first HFR Petten U 3 O 8 fuel elements are in progress. The full size ORR U 3 Si 2 fuel elements with so far highest uranium density of 4,76 g U/cm 3 meat achieved a burnup of 50 % loss of 235 U up to May 1983. One element was withdrawn from the reactor for PIE, the second will be irradiated to a burnup of 75 % loss of 235 U. The further development is concentrated on Usub(x)Sisub(y) fuel with highest uranium density. U 3 Si miniplates with up to 6,1 g U/cm 3 meat are supplied meeting the required specification, U 3 Si miniplates with 6,7 g U/cm 3 are in fabrication. (author)

Design, Manufacturing and Irradiation Behaviour of Fast Reactor Fuel. Proceedings of a Technical Meeting

International Nuclear Information System (INIS)

2013-04-01

Fast reactors are vital for ensuring the sustainability of nuclear energy in the long term. They offer vastly more efficient use of uranium resources and the ability to burn actinides, which are otherwise the long-lived component of high level nuclear waste. These reactors require development, qualification, testing and deployment of improved and innovative nuclear fuel and structural materials having very high radiation resistance, corrosion/erosion and other key operational properties. Several IAEA Member States have made efforts to advance the design and manufacture of technologies of fast reactor fuels, as well as to investigate their irradiation behaviour. Due to the acute shortage of fast neutron testing and post-irradiation examination facilities and the insufficient understanding of high dose radiation effects, there is a need for international exchange of knowledge and experience, generation of currently missing basic data, identification of relevant mechanisms of materials degradation and development of appropriate models. Considering the important role of nuclear fuels in fast reactor operation, the IAEA Technical Working Group on Fuel Performance and Technology (TWGFPT) proposed a Technical Meeting (TM) on 'Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels', which was hosted by the Institute of Physics and Power Engineering (IPPE) in Obninsk, Russian Federation, from 30 May to 3 June 2011. The TM included a technical visit to the fuel production plant MSZ in Elektrostal. The purpose of the meeting was to provide a forum to share knowledge, practical experience and information on the improvement and innovation of fuels for fast reactors through scientific presentations and brainstorming discussions. The meeting brought together 34 specialists from national nuclear agencies, R and D and design institutes, fuel vendors and utilities from 10 countries. The presentations were structured into four sections: R and D Programmes on FR Fuel

Modeling Thermal and Stress Behavior of the Fuel-clad Interface in Monolithic Fuel Mini-plates

International Nuclear Information System (INIS)

Miller, Gregory K.; Medvedev, Pavel G.; Burkes, Douglas E.; Wachs, Daniel M.

2010-01-01

As part of the Global Threat Reduction Initiative, a fuel development and qualification program is in process with the objective of qualifying very high density low enriched uranium fuel that will enable the conversion of high performance research reactors with operational requirements beyond those supported with currently available low enriched uranium fuels. The high density of the fuel is achieved by replacing the fuel meat with a single monolithic low enriched uranium-molybdenum fuel foil. Doing so creates differences in the mechanical and structural characteristics of the fuel plate because of the planar interface created by the fuel foil and cladding. Furthermore, the monolithic fuel meat will dominate the structural properties of the fuel plate rather than the aluminum matrix, which is characteristic of dispersion fuel types. Understanding the integrity and behavior of the fuel-clad interface during irradiation is of great importance for qualification of the new fuel, but can be somewhat challenging to determine with a single technique. Efforts aimed at addressing this problem are underway within the fuel development and qualification program, comprised of modeling, as-fabricated plate characterization, and post-irradiation examination. An initial finite element analysis model has been developed to investigate worst-case scenarios for the basic monolithic fuel plate structure, using typical mini-plate irradiation conditions in the Advanced Test Reactor. Initial analysis shows that the stress normal to the fuel-clad interface dominates during irradiation, and that the presence of small, rounded delaminations at the interface is not of great concern. However, larger and/or fuel-clad delaminations with sharp corners can create areas of concern, as maximum principal cladding stress, strain, displacement, and peak fuel temperature are all significantly increased. Furthermore, stresses resulting from temperature gradients that cause the plate to bow or buckle in

Guidebook on non-destructive examination of water reactor fuel

International Nuclear Information System (INIS)

1991-01-01

To date, a significant quantity of data has been collected and published on power reactor fuel examination to determine the performance when subjected to radiation. The data have been published in technical reports and papers in technical journals. However, the usefulness of the published data to the IAEA Member States is limited. This is due to a number of reasons, including the large variety of examination methods, incomplete documentation of the data and lack of sufficiently detailed information on pre-irradiation data and irradiation history. To alleviate some of these problems, the Agency initiated a Co-ordinated Research Programme in 1983 entitled ''Examination and Documentation Methodology for Water Reactor Fuel''. The programme meetings usually involved technical contributions from the programme participants, followed by a detailed discussion of the various examination methods presented in these contributions. Based on these discussions and contributions, a guidebook on the examination and documentation methodology for light and heavy water reactor fuel has been prepared. The guidebook addresses the most commonly used examination methods for the various water reactor fuel systems. Limitations of each of the measurement techniques are also discussed, including their accuracy and precision. A detailed description of the measurement equipment is given and the common methods of documenting the data are also addressed. With the adoption of the uniform set of procedures and documentation methods, it is hoped that the IAEA Member States will be able to use effectively both the existing data and the future data from the various national programmes. It is also expected that this guidebook will be useful for adaptation of measurement techniques that are unique to specific fuel systems to other fuel types. 59 refs, 33 figs, 4 tabs

Behaviour of irradiated uranium silicide fuel revisited

International Nuclear Information System (INIS)

Finlay, M. Ross; Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

2002-01-01

Irradiated U 3 Si 2 dispersion fuels demonstrate very low levels of swelling, even at extremely high burn-up. This behaviour is attributed to the stability of fission gas bubbles that develop during irradiation. The bubbles remain uniformly distributed throughout the fuel and show no obvious signs of coalescence. Close examination of high burn-up samples during the U 3 Si 2 qualification program revealed a bimodal distribution of fission gas bubbles. Those observations suggested that an underlying microstructure was responsible for the behaviour. An irradiation induced recrystallisation model was developed that relied on the presence of sufficient grain boundary surface to trap and pin fission gas bubbles and prevent coalescence. However, more recent work has revealed that the U 3 Si 2 becomes amorphous almost instantaneously upon irradiation. Consequently, the recrystallisation model does not adequately explain the nucleation and growth of fission gas bubbles in U 3 Si 2 . Whilst it appears to work well within the range of measured data, it cannot be relied on to extrapolate beyond that range since it is not mechanistically valid. A review of the mini-plates irradiated in the Oak Ridge Research Reactor from the U 3 Si 2 qualification program has been performed. This has yielded a new understanding of U 3 Si 2 behaviour under irradiation. (author)