WorldWideScience

Sample records for mox fuel non-destructive

  1. Transport of MOX fuel

    International Nuclear Information System (INIS)

    Porter, I.R.; Carr, M.

    1997-01-01

    The regulatory framework which governs the transport of MOX fuel is set out, including packages, transport modes and security requirements. Technical requirements for the packages are reviewed and BNFL's experience in plutonium and MOX fuel transport is described. The safety of such operations and the public perception of safety are described and the question of gaining public acceptance for MOX fuel transport is addressed. The paper concludes by emphasising the need for proactive programmes to improve the public acceptance of these operations. (Author)

  2. Mox fuels recycling

    International Nuclear Information System (INIS)

    Gay, A.

    1998-01-01

    This paper will firstly emphasis that the first recycling of plutonium is already an industrial reality in France thanks to the high degree of performance of La Hague and MELOX COGEMA's plants. Secondly, recycling of spent Mixed OXide fuel, as a complete MOX fuel cycle, will be demonstrated through the ability of the existing plants and services which have been designed to proceed with such fuels. Each step of the MOX fuel cycle concept will be presented: transportation, reception and storage at La Hague and steps of spent MOX fuel reprocessing. (author)

  3. BWRs with MOx fuel

    International Nuclear Information System (INIS)

    Demaziere, C.

    1999-01-01

    Calculations has been performed for loading BWRs with pure MOx or UOx/MOx fuel. It seems to be possible to load MOx bundles in BWRs, since most of the core characteristics are comparable with the ones of a full UOx core. Nevertheless two main problems arise: The shutdown margin at BOC is lower than 1%, this requires to have a new design for the control rods in order to increase their efficiency - but the problem can also be solved by modifying the Pu quality. The cores with MOx fuel are slightly less stable, unfortunately the simple model applied does not allow giving an absolute value for the decay ratio but only allows comparing the stability with the full UOx core

  4. MOX fuel reprocessing and recycling

    International Nuclear Information System (INIS)

    Guillet, J.L.

    1990-01-01

    This paper is devoted to the reprocessing of MOX fuel in UP2-800 plant at La Hague, and to the MOX successive reprocessing and recycling. 1. MOX fuel reprocessing. In a first step, the necessary modifications in UP2-800 to reprocess MOX fuel are set out. Early in the UP2-800 project, actions have been taken to reprocess MOX fuel without penalty. They consist in measures regarding: Dissolution; Radiological shieldings; Nuclear instrumentation; Criticality. 2. Mox successive reprocessing and recycling. The plutonium recycling in the LWR is now a reality and, as said before, the MOX fuel reprocessing is possible in UP2-800 plant at La Hague. The following actions in this field consist in verifying the MOX successive reprocessing and recycling possibilities. After irradiation, the fissile plutonium content of irradiated MOX fuel is decreased and, in this case, the re-use of plutonium in the LWR need an important increase of initial Pu enrichment inconsistent with the Safety reactor constraints. Cogema opted for reprocessing irradiated MOX fuel in dilution with the standard UO2 fuel in appropriate proportions (1 MOX for 4 UO2 fuel for instance) in order to save a fissile plutonium content compatible with MOX successive recycling (at least 3 recyclings) in LWR. (author). 2 figs

  5. MOX fuel fabrication technology in J-MOX

    International Nuclear Information System (INIS)

    Osaka, Shuichi; Yoshida, Ryouichi; Yamazaki, Yukiko; Ikeda, Hiroyuki

    2014-01-01

    Japan Nuclear Fuel Ltd. (JNFL) has constructed JNFL MOX Fuel Fabrication Plant (J-MOX) since 2010. The MIMAS process has been introduced in the powder mixing process from AREVA NC considering a lot of MOX fuel fabrication experiences at MELOX plant in France. The feed material of Pu for J-MOX is MH-MOX powder from Rokkasho Reprocessing Plant (RRP) in Japan. The compatibility of the MH-MOX powder with the MIMAS process was positively evaluated and confirmed in our previous study. This paper describes the influences of the UO2 powder and the recycled scrap powder on the MOX pellet density. (author)

  6. Development of MOX fuel database

    International Nuclear Information System (INIS)

    Ikusawa, Yoshihisa; Ozawa, Takayuki

    2007-03-01

    We developed MOX Fuel Database, which included valuable data from several irradiation tests in FUGEN and Halden reactor, for help of LWR MOX use. This database includes the data of fabrication and irradiation, and the results of post-irradiation examinations for seven fuel assemblies, i.e. P06, P2R, E03, E06, E07, E08 and E09, irradiated in FUGEN. The highest pellet peak burn-up reached ∼48GWd/t in MOX fuels, of which the maximum plutonium content was ∼6 wt%, irradiated in E09 fuel assembly without any failure. Also the data from the instrumented MOX fuels irradiated in HBWR to study the irradiation behavior of BWR MOX fuels under the steady state condition (IFA-514/565 and IFA-529), under the load-follow operation condition (IFA-554/555) and under the transit condition (IFA-591) are included in this database. The highest assembly burn-up reached ∼56 GWd/t in IFA-565 steady state irradiation test, and the maximum linear power of MOX fuel rods was 58.3-68.4 kW/m without any failure in IFA-591 ramp test. In addition, valuable instrument data, i.e. cladding elongation, fuel stack elongation, fuel center temperature and rod inner pressure were obtained from IFA-554/555 load-follow test. (author)

  7. MOX fuel fabrication at AECL

    International Nuclear Information System (INIS)

    Dimayuga, F.C.; Jeffs, A.T.

    1995-01-01

    Atomic Energy of Canada Limited's mixed-oxide (MOX) fuel fabrication activities are conducted in the Recycle Fuel Fabrication Laboratories (RFFL) at the Chalk River Laboratories. The RFFL facility is designed to produce experimental quantities of CANDU MOX fuel for reactor physics tests or demonstration irradiations. From 1979 to 1987, several MOX fuel fabrication campaigns were run in the RFFL, producing various quantities of fuel with different compositions. About 150 bundles, containing over three tonnes of MOX, were fabricated in the RFFL before operations in the facility were suspended. In late 1987, the RFFL was placed in a state of active standby, a condition where no fuel fabrication activities are conducted, but the monitoring and ventilation systems in the facility are maintained. Currently, a project to rehabilitate the RFFL and resume MOX fuel fabrication is nearing completion. This project is funded by the CANDU Owners' Group (COG). The initial fabrication campaign will consist of the production of thirty-eight 37-element (U,Pu)O 2 bundles containing 0.2 wt% Pu in Heavy Element (H.E.) destined for physics tests in the zero-power ZED-2 reactor. An overview of the Rehabilitation Project will be given. (author)

  8. High burnup MOX fuel assembly

    International Nuclear Information System (INIS)

    Blanpain, P.; Brunel, L.

    1999-01-01

    From the outset, the MOX product was required to have the same performance as UO 2 in terms of burnup and operational flexibility. In fact during the first years the UO 2 managements could not be applied to MOX. The changeover to an AFA 2G type fuel allowed an improvement in NPP operational flexibility. The move to the AFA 3G design fuel will enable an increase in the burnup of the MOX assemblies to the level of the UO 2 ones ('MOX Parity' project). But the FRAMATOME fuel development objective does not stop at the obtaining of parity between the current MOX and UO 2 products: this parity must remain guaranteed and the MOX managements must evolve in the same way as the UO 2 managements. The goal of the MOX product development programmes underway with COGEMA and the CEA is the demonstration over the next 10 years of a fuel capable of reaching burnups of 70 GWD/T. The research programmes focus on the fission gas release aspect, with three issues explored: optimization of pellet microstructures and validation in experimental reactor ; build-up of experience feedback from fission gas release at elevated burnups in commercial reactors, both for current and experimental products; adaptation and qualification of the design models and tools, over the ranges and for the products concerned. The product arising from these development programmes should be offered on the market around 2010. While meeting safety requirements, it will cater for the needs of the utilities in terms of product reliability, personnel dosimetry and kWh output costs (increase in burnup, NPP maneuverability and availability, minimization of process waste). (authors)

  9. The transportation of PuO2 and MOX fuel and management of irradiated MOX fuel

    International Nuclear Information System (INIS)

    Dyck, H.P.; Rawl, R.; Durpel, L. van den

    2000-01-01

    Information is given on the transportation of PuO 2 and mixed-oxide (MOX) fuel, the regulatory requirements for transportation, the packages used and the security provisions for transports. The experience with and management of irradiated MOX fuel and the reprocessing of MOX fuel are described. Information on the amount of MOX fuel irradiated is provided. (author)

  10. Public acceptance of MOX - fuel

    International Nuclear Information System (INIS)

    Huettmann, A.; Reddehase, C.G.

    1995-01-01

    In the Federal Republic of Germany 'Plutonium-Business' got fresh nutrient because of the carried out licensing of the use of Mixed Oxide (MOX)-fuel LWR and in connection with the negative attitude of the Hessian authorities, who are responsible for the licensing procedures of the production of MOX-fuel in the Siemens-factories at Hanau. The opponents of the peaceful use of nuclear energy try with the emotive expression 'Plutonium' (Pu) a frontal attack against the use of nuclear energy in Germany. They justify their actions with so-called safety deficits of the plants and increased danger of cancer in case of using MOX-fuel. (orig./HP)

  11. Experimental microstructures MOX fuels elaboration

    International Nuclear Information System (INIS)

    Gotta, M.J.; Dubois, S.; Lechelle, J.; Sornay, P.

    2000-01-01

    In order to propose a new MOX fuel, owning higher combustion rate, studies are realized at the CEA in collaboration with Cogema, EDF and Framatome. New microstructures of MOX are looked for around two approaches: the grains size and the plutonium distribution. These approaches are presented and discussed in this paper. The first one develops big grains microstructures obtained, either with anionic (sulfur), or cationic (Cr 2 O 3 ) additives. The second one concerns the CER-CER type composite microstructures. (A.L.B.)

  12. MOX fuel design and development consideration

    International Nuclear Information System (INIS)

    Yamate, K.; Abeta, S.; Suzuki, K.; Doi, S.

    1997-01-01

    Pu thermal utilization in Japan will be realized in several plants in late 1990's, and will be expanded gradually. For this target, adequacy of methods for MOX fuel design, nuclear design, and safety analysis has been evaluated by the committee of competent authorities organized by government in advance of the licensing application. There is no big difference of physical properties and irradiation behaviors between MOX fuel and UO 2 fuel, because Pu content of MOX fuel for Pu thermal utilization is low. The fuel design code for UO 2 fuel will be applied with some modifications, taking into account of characteristic of MOX fuel. For nuclear design, new code system is to be applied to treat the heterogeneity in MOX fuel assembly and the neutron spectrum interaction with UO 2 fuel more accurately. For 1/3 MOX fueled core in three loop plant, it was confirmed that the fuel rod mechanical design could meet the design criteria, with slight reduction of initial back-fitting pressure, and with appropriate fuel loading patterns in the core to match power with UO 2 fuel. With the increase of MOX fuel fraction in the core, control rod worth and boron worth decrease. Compensating the decrease by adding control rod and utilizing enriched B-10 in safety injection system, 100% MOX fueled core could be possible. Up to 1/3 MOX fueled core in three loop plant, no such modifications of the plant is necessary. The fraction of MOX fuel in PWR is designed to less than 1/3 in the present program. In order to improve Pu thermal utilization in future, various R and D program on fuel design and nuclear design are being performed, such as the irradiation program of MOX fuel manufactured through new process to the extent of high burnup. (author). 8 refs, 9 figs, 2 tabs

  13. Program on MOX fuel utilization in light water reactors

    International Nuclear Information System (INIS)

    Kenda, Hirofumi

    2000-01-01

    MOX fuel utilization program by the Japanese electric power companies was released in February, 1997. Principal philosophy for MOX fuel design is that MOX fuel shall be compatible with Uranium fuel and behavior of core loaded with MOX fuel shall be similar to that of conventional core. MOX fuel is designed so that geometry and nuclear capability of MOX fuel are equivalent to Uranium fuel. (author)

  14. MOX fuel for Indian nuclear power programme

    International Nuclear Information System (INIS)

    Kamath, H.S.; Anantharaman, K.; Purushotham, D.S.C.

    2000-01-01

    A sound energy policy and a sound environmental policy calls for utilisation of plutonium (Pu) in nuclear power reactors. The paper discusses the use of Pu in the form of mixed oxide (MOX) fuel in two Indian boiling water reactors (BWRs) at Tarapur. An industrial scale MOX fuel fabrication plant is presently operational at Tarapur which is capable of manufacturing MOX fuels for BWRs and in future for PHWRs. The plant can also manufacture mixed oxide fuel for prototype fast breeder reactor (PFBR) and development work in this regard has already started. The paper describes the MOX fuel manufacturing technology and quality control techniques presently in use at the plant. The irradiation experience of the lead MOX assemblies in BWRs is also briefly discussed. The key areas of interest for future developments in MOX fuel fabrication technology and Pu utilisation are identified. (author)

  15. MOX fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Shimada, Hidemitsu; Koyama, Jun-ichi; Aoyama, Motoo

    1998-01-01

    The MOX fuel assembly of the present invention is of a c-lattice type loaded to a BWR type reactor. 74 MOX fuel rods filled with mixed oxides of uranium and plutonium and two water rods disposed to a space equal to that for 7 MOX fuel rods are arranged in 9 x 9 matrix. MOX fuel rods having the lowest enrichment degree are disposed to four corners of the 9 x 9 matrix. The enrichment degree means a ratio of the weight of fission products based on the total weight of fuels. Two MOX fuel rods having the same enrichment degree are arranged in each direction so as to be continuous from the MOX fuel rods at four corners in the direction of the same row and different column and same column and the different row. In addition, among the outermost circumferential portion of the 9 x 9 matrix, MOX fuel rods having a lower enrichment degree next to the MOX fuel rods having the lowest enrichment degree are arranged, each by three to a portion where MOX fuel rods having the lowest enrichment degree are not disposed. (I.N.)

  16. Fission gas release behaviour in MOX fuels

    International Nuclear Information System (INIS)

    Viswanathan, U.K.; Anantharaman, S.; Sahoo, K.C.

    2002-01-01

    As a part of plutonium recycling programme MOX (U,Pu)O 2 fuels will be used in Indian boiling water reactors (BWR) and pressurised heavy water reactors (PHWR). Based on successful test irradiation of MOX fuel in CIRUS reactor, 10 MOX fuel assemblies have been loaded in the BWR of Tarapur Atomic Power Station (TAPS). Some of these MOX fuel assemblies have successfully completed the initial target average burnup of ∼16,000 MWD/T. Enhancing the burnup target of the MOX fuels and increasing loading of MOX fuels in TAPS core will depend on the feedback information generated from the measurement of released fission gases. Fission gas release behaviour has been studied in the experimental MOX fuel elements (UO 2 - 4% PuO 2 ) irradiated in pressurised water loop (PWL) of CIRUS. Eight (8) MOX fuel elements irradiated to an average burnup of ∼16,000 MWD/T have been examined. Some of these fuel elements contained controlled porosity pellets and chamfered pellets. This paper presents the design details of the experimental set up for studying fission gas release behaviour including measurement of gas pressure, void volume and gas composition. The experimental data generated is compared with the prediction of fuel performance modeling codes of PROFESS and GAPCON THERMAL-3. (author)

  17. A MOX fuel attribute monitor

    International Nuclear Information System (INIS)

    Bliss, Mary; Jordan, David V.; Barnett, Debra S.; Redding, Rebecca L.; Pearce, Stephen K.

    2007-01-01

    Euratom performs safeguards monitoring of Fresh MOX fuel for domestic power production in the European Union. Video cameras monitor the reactor storage ponds. If video surveillance is lost for a certain amount of time a measurement is required to verify that no fuel was diverted. The attribute measurement to verify the continued presence of MOX fuel is neutron emission. Ideally this measurement would be made without moving or handling the fuel rod assembly. A prototype attribute measurement system was made using scintillating neutron sensitive glass waveguides developed by Pacific Northwest National Laboratory. Short lengths (5-20 cm) of the neutron sensitive fiber were mechanically spliced to 15 m lengths of commercial high numerical aperture fiber optic cable (Ceramoptec Optran Ultra 0.44). The light detector is a Hamamatsu R7400P photomultiplier tube. An electronics package was built to use the sensors with a GBS Elektronik MCA-166 multichannel analyzer and user interface. The MCA-166 is the system most commonly used by Euratom inspectors. It can also be run from a laptop computer using Maestro (Ortec) or other software. A MCNP model was made to compare to measurements made with several neutron sources including NIST traceable 252 Cf

  18. Advanced analysis technology for MOX fuel

    International Nuclear Information System (INIS)

    Hiyama, T.; Kamimura, K.

    1997-01-01

    PNC has developed MOX fuels for advanced thermal reactor (ATR) and fast breeder reactor (FBR). The MOX samples have been chemically analysed to characterize the MOX fuel for JOYO, MONJU, FUGEN and so on. The analysis of the MOX samples in glove box has required complicated and highly skilled operations. Therefore, for quality control analysis of the MOX fuel in a fabrication plant, simple, rapid and accurate analysis methods are necessary. To solve the above problems instrumental analysis and techniques were developed. This paper describes some of the recent developments in PNC. 2. Outline of recently developed analysis methods by PNC. 2.1 Determination of oxygen to metal atomic ratio (O/M) in MOX by non-dispersive infrared spectrophotometry after inert gas fusion. 7 refs, 9 figs, 4 tabs

  19. Plant overview of JNFL MOX fuel fabrication plant (J-MOX)

    International Nuclear Information System (INIS)

    Hiruta, Kazuhiko; Suzuki, Masataka; Shimizu, Junji; Suzuki, Kazumi; Yamamoto, Yutaka; Deguchi, Morimoto; Fujimaki, Kazunori

    2005-01-01

    In April 2005, JNFL submitted METI an application for the permission of MOX fuel fabrication business for JNFL MOX Fuel Fabrication Plant (J-MOX). Accordingly, safeguards formalities and discussion with the Agency have been also started for J-MOX as an official project. This report describes J-MOX plant overview and also presents outline of J-MOX by focusing on safeguards features and planned material accountancy method. (author)

  20. Non-destructive analysis of spent nuclear fuel

    International Nuclear Information System (INIS)

    Popovic, D.

    1961-12-01

    Nondestructive analysis of fuel elements dealt with determining the isotope contents which provide information about the burnup level, quantities of fission products and neutron-multiplication properties of the irradiated fuel. Methods for determination of the isotope ratio of the spent fuel are both numerical and experimental. This report deals with the experimental method. This means development of the experimental methods for direct measurement of the isotope content. A number of procedures are described: measurements of α, β and γ activities of the isotopes; measurement of secondary effects of nuclear reactions with thermal neutrons and fast neutrons; measurement of cross sections; detection of prompt and delayed neutrons

  1. Recycling of MOX fuel for LWRs

    International Nuclear Information System (INIS)

    Joo, Hyung Kook; Oh, Soo Youl

    1992-01-01

    The status and issues related to the thermal recycling of reprocessed nuclear fuels have been reviewed. It is focused on the use of reprecessed plutonium in the form of mixed oxide (MOX) for a light water reactor and the review on reprocessing and fabrication processes is beyond the scope. In spite of the difference in the nuclear characteristics between plutonium and uranium isotopes, the neutronics behavior in a core with MOX fuels is similar to that with normal uranium fuels. However, since the neutron spectrum is hardened in a core with MOX, the Doppler, viod, and moderator temperature coefficients become more negative and the control rod and boron worths are slightly reduced. Therefore, the safety will be evaluated carefully in addition to the core neutronics analysis. The MOX fuel rod behavior related to the rod performance such as the pellet to clad interaction and fission gas release is also similar to that of uranium rods, and no specific problem arises. Substituting MOX fuels for a portion of uranium fuels, it is estimated that the savings be about 25% in uranium ore and 10% in uranium enrichment service requirements. The use of MOX fuel in LWRs has been commercialized in European countries including Germany, France, Belgium, etc., and a demonstration program has been pursued in Japan for the commercial utilization in the late 1990s. Such a worldwide trend indicates that the utilization of MOX fuel in LWRs is a proven technology and meets economics criteria. (Author)

  2. 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)

  3. 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

  4. Non-destructive delamination detection in solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Gazzarri, J.I.; Kesler, O. [Department of Mechanical Engineering, University of British Columbia, 2054-6250 Applied Science Lane, Vancouver, BC V6T 1Z4 (Canada)

    2007-05-15

    A finite element model has been developed to simulate the steady state and impedance behaviour of a single operating solid oxide fuel cell (SOFC). The model results suggest that electrode delamination can be detected minimally-invasively by using electrochemical impedance spectroscopy. The presence of cathode delamination causes changes in the cell impedance spectrum that are characteristic of this type of degradation mechanism. These changes include the simultaneous increase in both the series and polarization resistances, in proportion to the delaminated area. Parametric studies show the dependence of these changes on the extent of delamination, on the operating point, and on the kinetic characteristics of the fuel cell under study. (author)

  5. Method and equipment for the non-destructive analysis of nuclear fuels

    International Nuclear Information System (INIS)

    Michaelis, W.

    1975-01-01

    This is a method for the non-destructive analysis of the content of fissile isotopes in nuclear fuels. In this analysis a neutron beam is directed to the nuclear fuel which is to be analysed. The beam penetrates the nuclear fuel, thus causing a secondany radiation by nuclear reactions which reaches a space directly surrounding the nuclear fuel and is measuned there. (orig./UA) [de

  6. MOX fuel transport: the French experience

    International Nuclear Information System (INIS)

    Sanchis, H.; Verdier, A.; Sanchis, H.

    1999-01-01

    In the back-end of the fuel cycle, several leading countries have chosen the Reprocessing, Conditioning, Recycling (RCR) option. Plutonium recycling in the form of MOX fuel is a mature industry, with successful operational experience and large-scale fabrication plants an several European countries. The COGEMA Group has developed the industrialized products to master the RCR operation including transport COGEMA subsidiary, TRANSNUCLEAIRE has been operating MOX fuel transports on an industrial scale for more than 10 years. In 1998, around 200 transports of Plutonium materials have been organised by TRANSNUCLEAIRE. These transports have been carried out by road between various facilities in Europe: reprocessing plants, manufacturing plants and power plants. The materials transported are either: PuO 2 and MOX powder; BWR and PWR MOX fuel rods; BWR and PWR MOX fuel assemblies. Because MOX fuel transport is subject to specific safety, security and fuel integrity requirements, the MOX fuel transport system implemented by TRANSNUCLEAIRE is fully dedicated. Packaging have been developed, licensed and manufactured for each kind of MOX material in compliance with relevant regulations. A fleet of vehicles qualified according to existing physical protection regulations is operated by TRANSNUCLEAIRE. TRANSNUCLEAIRE has gained a broad experience in MOX transport in 10 years. Technical and operational know-how has been developed and improved for each step: vehicles and packaging design and qualification; vehicle and packaging maintenance; transport operations. Further developments are underway to increase the payload of the packaging and to improve the transport conditions, safety and security remaining of course top priority. (authors)

  7. The MOX fuel in France

    International Nuclear Information System (INIS)

    2011-01-01

    This document briefly describes the MOX production cycle which is performed in the MELOX plant in Marcoule by AREVA. It briefly indicates the main risks occurring during the whole MOX production and use cycle. They are associated with MOX production (high neutron and gamma dose rates, contamination, criticality, heat release), transportation, its use in reactors, its storage in pools after irradiation. All these stages need radiation protection measures

  8. Characteristics of plutonium, curium and uranium in hulls of FUGEN MOX spent fuel by destructive analysis

    International Nuclear Information System (INIS)

    Iijima, Shizuka; Goto, Yuichi; Samoto, Hirotaka; Shichi, Ryo; Shimizu, Takenori

    2011-01-01

    We have been developing a non-destructive assay system called hulls monitor for nuclear fuel materials retained in hulls at the Tokai Reprocessing Plant (TRP). The hulls monitor is based on a passive neutron measurement method, so its applicability should be evaluated by a destructive analysis of hulls that are recovered from the reprocessing process. In this study, hulls came from the Advanced Thermal Reactor (ATR) FUGEN were taken from the dissolution process of TRP and destructively analyzed. Two kinds of hulls from ATR-MOX spent fuel assemblies and from ATR-UO 2 spent fuel assemblies were taken and soaked with nitric acid then fused with ammonium hydrogen sulfate, followed by Pu, 244 Cm, U mass determination by alpha spectrometry and ICP-AES. The characteristics of hulls came from MOX spent fuel assemblies were revealed by comparison of ATR-MOX spent fuel with ATR-UO 2 spent fuel. (author)

  9. Non-destructive control of cladding thickness of fuel elements for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karlov, Y.; Zhukov, Y.; Chashchin, S

    1997-07-01

    The control method of fuel elements for research reactors by means of measuring beta particles back scattering made it possible to perform complete automatic non-destructive control of internal and external claddings at our plant. This control gives high guarantees of the fuel element correspondence to the requirements. The method can be used to control the three-layer items of different geometry, including plates. (author)

  10. Fuel production for LWRs - MOX fuel aspects

    International Nuclear Information System (INIS)

    Deramaix, P.

    2005-01-01

    Plutonium recycling in Light Water Reactors is today an industrial reality. It is recycled in the form of (U, Pu)O 2 fuel pellets (MOX), fabricated to a large extent according to UO 2 technology and pellet design. The similarity of physical, chemical, and neutron properties of both fuels also allows MOX fuel to be burnt in nuclear plants originally designed to burn UO 2 . The industrial processes presently in use or planned are all based on a mechanical blending of UO 2 and PuO 2 powders. To obtain finely dispersed plutonium and to prevent high local concentration of plutonium, the feed materials are micronised. In the BNFL process, the whole (UO 2 , PuO 2 ) blend is micronised by attrition milling. According to the MIMAS process, developed by BELGONUCLEAIRE, a primary blend made of UO 2 containing about 30% PuO 2 is micronised in a ball mill, afterwards this primary blend is mechanically diluted in UO 2 to obtain the specified Pu content. After mixing, the (U, Pu)O 2 powder is pressed and the pellets are sintered. The sintering cover gas contains moisture and 5 v/o H 2 . Moisture increases the sintering process and the U-Pu interdiffusion. After sintering and grinding, the pellets are submitted to severe controls to verify conformity with customer specifications (fissile content, Pu distribution, surface condition, chemical purity, density, microstructure). (author)

  11. MOX fuel fabrication: Technical and industrial developments

    International Nuclear Information System (INIS)

    Lebastard, G.; Bairiot, H.

    1990-01-01

    The plutonium available in the near future is generally estimated rather precisely on the basis of the reprocessing contracts and the performance of the reprocessing plants. A few years ago, decision makers were convinced that a significant share of this fissile material would be used as the feed material for fast breeder reactors (FBRs) or other advanced reactors. The facts today are that large reprocessing plants are coming into commercial operations: UP3 and soon UP2-800 and THORP, but that FBR deployment is delayed worldwide. As a consequence, large quantities of plutonium will be recycled in light water reactors as mixed oxide (MOX) fuels. MOX fuel technology has been properly demonstrated in the past 25 years. All specific problems have been addressed, efficient fabrication processes and engineering background have been implemented to a level of maturity which makes MOX fuel behaving as well as Uranium fuel. The paper concentrates on todays MOX fabrication expertise and presents the technical and industrial developments prepared by the MOX fuel fabrication industry for this last decade of the century

  12. Design of the MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Johnson, J.V.; Brabazon, E.J.

    2001-01-01

    A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)

  13. Design of the MOX fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.V. [MFFF Technical Manager, U.S. dept. of Energy, Washington, DC (United States); Brabazon, E.J. [MFFF Engineering Manager, Duke Cogema Stone and Webster, Charlotte, NC (United States)

    2001-07-01

    A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)

  14. Overview of MOX fuel fabrication achievements

    International Nuclear Information System (INIS)

    Bairiot, H.; Vliet, J. van; Chiarelli, G.; Edwards, J.; Nagai, S.H.; Reshetnikov, F.

    2000-01-01

    Such overview having been adequately covered in an OECD/NEA publication providing the situation as of end 1994, this paper is mainly devoted to an update as of end 1998. The Belgian plant, Belgonucleaire/Dessel, is now dedicated exclusively to the fabrication of MOX fuel and has operated consistently around its nameplate capacity (35tHM/a) through the 1990s involving a large variety of PWR and BWR fuels. The two French plants have also achieved routine operation during the 1990s. CFCa, historically the largest FBR MOX fuel manufacturer, is utilizing the genuine COCA process for that type of fuel and the MIMAS process for LWR fuel: a nominal capacity (40 tHM/a) has been gradually approached. MELOX has operated at 100 tHM/a, as defined in the operating licence granted originally. The British plant, MDF/Sellafield with 8tHM/a nameplate capacity is devoted to fuel and has manufactured several small fabrication campaigns. In Japan, JNC operates three facilities located at Tokai: PFDF, devoted to basic research and fabrication of test fuels, PFFF/ATR line, for the fabrication of Fugen fuel and of corresponding fuel for the critical facility DCA, and PFPF for the fabrication of FBR fuel. In Russia, fabrication techniques have been developed to fuel four BN-800 FBRs contemplated to be constructed and be fuelled with the civilian Pu stockpile. Two demonstration facilities Paket (Mayak) and RIAR (Dimitrovgrad) fabricated respectively pellet and vipac type FBR MOX fuel for BR-5, BOR-60, BN-350 and BN-600. The paper includes a brief description of each of the fabrication routes mentioned, as well as the production of respectively LWR and FBR MOX fuel in each fabrication facility, since the start-up of the plant, since 1 January 1993 and since 1 January 1998 up to 31 December 1998. (author)

  15. Non destructive examination of UN / U-Si fuel pellets using neutrons (preliminary assessment)

    Energy Technology Data Exchange (ETDEWEB)

    Bourke, Mark Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Voit, Stewart Lancaster [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Losko, Adrian S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tremsin, Anton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-31

    Tomographic imaging and diffraction measurements were performed on nine pellets; four UN/ U Si composite formulations (two enrichment levels), three pure U3Si5 reference formulations (two enrichment levels) and two reject pellets with visible flaws (to qualify the technique). The U-235 enrichments ranged from 0.2 to 8.8 wt.%. The nitride/silicide composites are candidate compositions for use as Accident Tolerant Fuel (ATF). The monophase U3Si5 material was included as a reference. Pellets from the same fabrication batches will be inserted in the Advanced Test Reactor at Idaho during 2016. The goal of the Advanced Non-destructive Fuel Examination work package is the development and application of non-destructive neutron imaging and scattering techniques to ceramic and metallic nuclear fuels. Data reported in this report were collected in the LANSCE run cycle that started in September 2015 and ended in March 2016. Data analysis is ongoing; thus, this report provides a preliminary review of the measurements and provides an overview of the characterized samples.

  16. Technology developments for Japanese BWR MOX fuel utilization

    International Nuclear Information System (INIS)

    Oguma, M.; Mochida, T.; Nomata, T.; Asahi, K.

    1997-01-01

    The Long-Term Program for Research, Development and Utilization of Nuclear Energy established by the Atomic Energy Commission of Japan asserts that Japan will promote systematic utilization of MOX fuel in LWRs. Based on this Japanese nuclear energy policy, we have been pushing development of MOX fuel technology aimed at future full scale utilization of this fuel in BWRs. In this paper, the main R and D topics are described from three subject areas, MOX core and fuel design, MOX fuel irradiation behaviour, and MOX fuel fabrication technology. For the first area, we explain the compatibility of MOX fuel with UO 2 core, the feasibility of the full MOX core, and the adaptability of MOX design methods based on a mock-up criticality experiment. In the second, we outline the Tsuruga MOX irradiation program and the DOMO program, and suggest that MOX fuel behaviour is comparable to ordinary BWR UO 2 fuel behaviour. In the third, we examine the development of a fully automated MOX bundle assembling apparatus and its features. (author). 14 refs, 11 figs, 3 tabs

  17. Performance evaluation of WDXRF as a process control technique for MOX fuel fabrication

    International Nuclear Information System (INIS)

    Pandey, A.; Khan, F.A.; Das, D.K.; Behere, P.G.; Afzal, Mohd

    2015-01-01

    This paper presents studies on Wavelength Dispersive X-Ray Fluorescence (WDXRF), as a powerful non destructive technique (NDT) for the compositional analysis of various types of MOX fuels. The sample has come after mixing and milling of UO 2 and PuO 2 powder for the estimation of plutonium, as a process control step of fabrication of (U, Pu)O 2 mixed oxide (MOX) fuel. For the characterization for heavy metal in various MOX fuel, a WDXRF method was established as a process control technique. The attractiveness of our system is that it can analyze the samples in solid form as well as in liquid form. The system is adapted in a glove box for handling of plutonium based fuels. The glove box adapted system was optimized with Uranium and Thorium based MOX sample before introduction of Pu. Uranium oxide and thorium oxide have been estimated in uranium thorium MOX samples. Standard deviation for the analysis of U 3 O 8 and ThO 2 were found to be 0.14 and 0.15 respectively. The results are validated against the conventional wet chemical methods of analysis. (author)

  18. Non-destructive methods of control of thermo-physical properties of fuel rods

    International Nuclear Information System (INIS)

    Kruglov, A B; Kruglov, V B; Kharitonov, V S; Struchalin, P G; Galkin, A G

    2017-01-01

    Information about the change of thermal properties of the fuel elements needed for a successful and safe operation of the nuclear power plant. At present, the existing amount of information on the fuel thermal conductivity change and “fuel-shell” thermal resistance is insufficient. Also, there is no technique that would allow for the measurement of these properties on the non-destructive way of irradiated fuel elements. We propose a method of measuring the thermal conductivity of the fuel in the fuel element and the contact thermal resistance between the fuel and the shell without damaging the integrity of the fuel element, which is based on laser flash method. The description of the experimental setup, implementing methodology, experiments scheme. The results of test experiments on mock-ups of the fuel elements and their comparison with reference data, as well as the results of numerical modeling of thermal processes that occur during the measurement. Displaying harmonization of numerical calculation with the experimental thermograms layout shell portions of the fuel cell, confirming the correctness of the calculation model. (paper)

  19. Determination of nuclear fuel burnup by non-destructive gamma spectroscopy

    International Nuclear Information System (INIS)

    Soares, A.J.

    1979-01-01

    The determination of nuclear fuel burnup by the non-destructive gamma spectroscopy method is studied. A MTR (Materials Testing Reactor) -type fuel element is used in the measurement. The fuel element was removed from the reactor core in 1958 and, because of the long decay time, show only one peak in is gamma spectrum at 661.6 Kev. Corresponding to 137 Cs. Measurements are made at 330 points of the element using a Nal detector and the final result revealed that the quantity of 235 U consumed was 3.3 +- 0,8 milligram in the entire element. The effect of the migration of 137 Cs in the element is neglected in view of the fact that it occurs only when the temperature is above 1000 0 C, which is not the case in IEAR-1. (Author)

  20. Operational experience in the non-destructive assay of fissile material in General Electric's nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Stewart, J.P.

    1976-01-01

    Operational experience in the non-destructive assay of fissile material in a variety of forms and containers and incorporation of the assay devices into the accountability measurement system for General Electric's Wilmington Fuel Fabrication Facility measurement control programme is detailed. Description of the purpose and related operational requirements of each non-destructive assay system is also included. In addition, the accountability data acquisition and processing system is described in relation to its interaction with the various non-destructive assay devices and scales used for accountability purposes within the facility. (author)

  1. Role of non-destructive examinations in leak testing of glove boxes for industrial scale plutonium handling at nuclear fuel fabrication facility along with case study

    International Nuclear Information System (INIS)

    Aher, Sachin

    2015-01-01

    Non Destructive Examinations has the prominent role at Nuclear Fuel Fabrication Facilities. Specifically NDE has contributed at utmost stratum in Leak Testing of Glove Boxes and qualifying them as a Class-I confinement for safe Plutonium handling at industrial scale. Advanced Fuel Fabrication Facility, BARC, Tarapur is engaged in fabrication of Plutonium based MOX (PuO 2 , DDUO 2 ) fuel with different enrichments for first core of PFBR reactor. Alpha- Leak Tight Glove Boxes along with HEPA Filters and dynamic ventilation form the promising engineering system for safe and reliable handling of plutonium bearing materials considering the radiotoxicity and risk associated with handling of plutonium. Leak Testing of Glove Boxes which involves the leak detection, leak rectification and leak quantifications is major challenging task. To accomplish this challenge, various Non Destructive Testing methods have assisted in promising way to achieve the stringent leak rate criterion for commissioning of Glove Box facilities for plutonium handling. This paper highlights the Role of various NDE techniques like Soap Solution Test, Argon Sniffer Test, Pressure Drop/Rise Test etc. in Glove Box Leak Testing along with procedure and methodology for effective rectification of leakage points. A Flow Chart consisting of Glove Box leak testing procedure starting from preliminary stage up to qualification stage along with a case study and observations are discussed in this paper. (author)

  2. Non-destructive evaluation methods to improve quality control in low enrichment MTR fuel plate production

    International Nuclear Information System (INIS)

    Milne, J.M.; Lidington, B.; Hawker, B.M.

    1991-01-01

    This paper summarises some preliminary non-destructive measurements made recently at the Harwell Laboratory on a prototype low enrichment MTR fuel plate. The measurements were intended to indicate the potential of two different techniques for improving quality control in plate production. Pulse Video Thermography (PVT) is being considered as an alternative to ultrasound transmission measurements for the detection and sizing of lack of thermal bonding between the fuel and the clad layers, either to verify the indications from the established ultrasonic methods before destroying the plate or as a replacement method of inspection. High frequency pulse-echo ultrasonics is being considered for providing maps of clad layer thickness on each side of the plate. The measurements have indicated the potential for both methods, but more work is required, using a test plate containing controlled defects, to establish their capability. (orig.)

  3. Characterization of spent fuel assemblies for storage facilities using non destructive assay

    International Nuclear Information System (INIS)

    Lebrun, A.; Bignan, G.; Recroix, H.; Huver, M.

    1999-01-01

    Many non destructive assay (NDA) techniques have been developed by the French Atomic Energy Commission (CEA) for spent fuel characterization and management. Passive and active neutron methods as well as gamma spectrometric methods have been carried out and applied to industrial devices like PYTHON TM and NAJA. Many existing NDA methods can be successfully applied to storage, but the most promising are the neutron methods combined with on line evolution codes. For dry storage applications, active neutron measurements require further R and D to achieve accurate results. Characterization data given by NDA instruments can now be linked to automatic fuel recognition. Both information can feed the storage management software in order to meet the storage operation requirements like: fissile mass inventory, operators declaration consistency or automatic selection of proper storage conditions. (author)

  4. Memento. Maritime transport of MOX fuels from Europe to Japan

    International Nuclear Information System (INIS)

    1999-07-01

    The maritime transport of MOX fuels from Europe to Japan represents the last of the 3 steps of transport of the nuclear fuel reprocessing-recycling program settled between ORC (Japan), BNFL (UK) and Cogema (France). This document summarizes the different aspects of this program: the companies concerned, the physical protection measures, the US-Japan agreements (accompanying warship), the in-depth safety, the handling of MOX fuels (containers and ships), and the Japan MOX fuel needs. (J.S.)

  5. Waste management in MOX fuel fabrication plants

    International Nuclear Information System (INIS)

    Schneider, V.

    1982-01-01

    After a short description of a MOX fuel fabrication plant's activities the waste arisings in such a plant are discussed according to nature, composition, Pu-content. Experience has shown that proper recording leads to a reduction of waste arisings by waste awareness. Aspects of the treatment of α-waste are given and a number of treatment processes are reviewed. Finally, the current waste management practice and the α-waste treatment facility under construction at ALKEM are outlined. (orig./RW)

  6. Models for MOX fuel behaviour. A selective review

    International Nuclear Information System (INIS)

    Massih, Ali R.

    2006-01-01

    This report reviews the basic physical properties of light water reactor mixed-oxide (MOX) fuel comprising nuclear characteristics, thermal properties such as melting temperature, thermal conductivity, thermal expansion, and heat capacity, and compares these with properties of conventional UO 2 fuel. These properties are generally well understood for MOX fuel and are well described by appropriate models developed for engineering analysis. Moreover, certain modelling approaches of MOX fuel in-reactor behaviour, regarding densification, swelling, fission product gas release, helium release, fuel creep and grain growth, are evaluated and compared with the models for UO 2 . In MOX fuel the presence of plutonium rich agglomerates adds to the complexity of fuel behaviour on the micro scale. In addition, we survey the recent fuel performance experience and post irradiation examinations on several types of MOX fuel types. We discuss the data from these examinations, regarding densification, swelling, fission product gas release and the evolution of the microstructure during irradiation. The results of our review indicate that in general MOX fuel has a higher fission gas release and helium release than UO 2 fuel. Part of this increase is due to the higher operating temperatures of MOX fuel relative to UO 2 fuel due to the lower thermal conductivity of MOX material. But this effect by itself seems to be insufficient to make for the difference in the observed fission gas release of UO 2 vs. MOX fuel. Furthermore, the irradiation induced creep rate of MOX fuel is higher than that of UO 2 . This effect can reduce the pellet-clad interaction intensity in fuel rods. Finally, we suggest that certain physical based approaches discussed in the report are implemented in the fuel performance code to account for the behaviour of MOX fuel during irradiation

  7. Models for MOX fuel behaviour. A selective review

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2006-12-15

    This report reviews the basic physical properties of light water reactor mixed-oxide (MOX) fuel comprising nuclear characteristics, thermal properties such as melting temperature, thermal conductivity, thermal expansion, and heat capacity, and compares these with properties of conventional UO{sub 2} fuel. These properties are generally well understood for MOX fuel and are well described by appropriate models developed for engineering analysis. Moreover, certain modelling approaches of MOX fuel in-reactor behaviour, regarding densification, swelling, fission product gas release, helium release, fuel creep and grain growth, are evaluated and compared with the models for UO{sub 2}. In MOX fuel the presence of plutonium rich agglomerates adds to the complexity of fuel behaviour on the micro scale. In addition, we survey the recent fuel performance experience and post irradiation examinations on several types of MOX fuel types. We discuss the data from these examinations, regarding densification, swelling, fission product gas release and the evolution of the microstructure during irradiation. The results of our review indicate that in general MOX fuel has a higher fission gas release and helium release than UO{sub 2} fuel. Part of this increase is due to the higher operating temperatures of MOX fuel relative to UO{sub 2} fuel due to the lower thermal conductivity of MOX material. But this effect by itself seems to be insufficient to make for the difference in the observed fission gas release of UO{sub 2} vs. MOX fuel. Furthermore, the irradiation induced creep rate of MOX fuel is higher than that of UO{sub 2}. This effect can reduce the pellet-clad interaction intensity in fuel rods. Finally, we suggest that certain physical based approaches discussed in the report are implemented in the fuel performance code to account for the behaviour of MOX fuel during irradiation.

  8. BNFL assessment of methods of attaining high burnup MOX fuel

    International Nuclear Information System (INIS)

    Brown, C.; Hesketh, K.W.; Palmer, I.D.

    1998-01-01

    It is clear that in order to maintain competitiveness with UO 2 fuel, the burnups achievable in MOX fuel must be enhanced beyond the levels attainable today. There are two aspects which require attention when studying methods of increased burnups - cladding integrity and fuel performance. Current irradiation experience indicates that one of the main performance issues for MOX fuel is fission gas retention. MOX, with its lower thermal conductivity, runs at higher temperatures than UO 2 fuel; this can result in enhanced fission gas release. This paper explores methods of effectively reducing gas release and thereby improving MOX burnup potential. (author)

  9. Transport of MOX fuel from Europe to Japan

    International Nuclear Information System (INIS)

    2002-01-01

    The MOX fuel transports from Europe to Japan represent a main part in the implementing of the Japan nuclear program. They complement the 160 transports of spent fuels realized from Japan to Europe and the vitrified residues return from France to Japan. In this framework the document presents the MOX fuel, the use of the MOX fuel in reactor, the proliferation risks, the MOX fuel transport to Japan, the public health, the transport regulations, the safety and the civil liability. (A.L.B.)

  10. RIA tests in CABRI with MOX fuel

    International Nuclear Information System (INIS)

    Schmitz, F.; Papin, J.; Gonnier, C.

    2000-01-01

    Three MOX-fuel tests have been successfully performed within the framework of the CABRI REP-Na test program. From the experimental findings which are presently available, no evidence for thermal effects resulting from the heterogeneous nature of the fuel can be given. There are very clear hints however that fission gas effects are enhanced with regard to the behaviour of UO 2 . The clad rupture observed in REP-Na 7 is of different nature than the failures observed in Cabri tests with UO 2 fuel. Failures of UO 2 fuel rods only occurred when the clad mechanical properties were severely affected by the presence of hydride blisters, while in REP-Na 7 a clear indication is made that the loading potential of the MOX fuel pellets was high enough to break a sound cladding. Concerning the transient fuel behaviour after reaching the critical heat-flux under reactor typical conditions (pressure, temperature and flow), no data base could be provided by the tests in the present sodium test loop (as for the UO 2 fuel behaviour). The IPSN project to implement into the Cabri reactor a pressurised water loop which will allow to simulate the complete RIA accident sequence under PWR reactor typical conditions, aims at providing this missing data base. (author)

  11. A versatile passive and active non-destructive device for spent fuel assemblies monitoring

    International Nuclear Information System (INIS)

    Berne, R.; Bignan, G.; Andrieu, G.; Dethan, B.

    1993-01-01

    The monitoring of spent fuel assemblies in reactor pools or in reprocessing plants with NDA methods is interesting (non-destructivity, non-intrusivity) for process control, safety-criticality and/or nuclear material management. In this context, the authors present the results of the development and design of a prototype device (physical methods used, qualification...) called PYTHON. The aim of PYTHON is to check the declared characteristic values of an irradiated assembly before taking it into a transport cask for safety criticality control. The PYTHON device consists of a detector head in two sections and a 252 Cf source if active neutron counting is to be used. Each section of the detection head consists of two detectors: one fission chamber and one ionization chamber

  12. Mox fuel experience: present status and future improvements

    International Nuclear Information System (INIS)

    Blanpain, P.; Chiarelli, G.

    2001-01-01

    Up to December 2000, more than 1700 MOX fuel assemblies have been delivered by Framatome ANP/Fragema to 20 French, 2 Belgian and 3 German PWRs. More than 1000 MOX fuel assemblies have been delivered by Framatome ANP GmbH (formerly Siemens) to 11 German PWRs and BWRs and to 3 Swiss PWRs. Operating MOX fuel up to discharge burnups of about 45,000 MWd/tM is done without any penalty on core operating conditions and fuel reliability. Performance data for fuel and materials have been obtained from an outstanding surveillance program. The examinations have concluded that there have been no significant differences in MOX fuel assembly characteristics relative to UO 2 fuel. The data from these examinations, combined with a comprehensive out-of-core and in-core analytical test program on the current fuel products, are being used to confirm and upgrade the design models necessary for the continuing improvement of the MOX product. As MOX fuel has reached a sufficient maturity level, the short term step is the achievement of the parity between UO 2 and MOX fuels in the EdF French reactors. This involves a single operating scheme for both fuels with an annual quarter core reload type and an assembly discharge burnup goal of 52,000 MWd/tM. That ''MOX parity'' product will use the AFA-3G assembly structure which will increase the fuel rod design margins with regards to the end-of-life internal pressure criteria. But the fuel development objective is not limited to the parity between the current MOX and UO 2 products: that parity must remain guaranteed and the MOX fuel managements must evolve in the same way as the UO 2 ones. The goal of the MOX product development program underway in France is the demonstration over the next ten years of a fuel capable of reaching assembly burnups of 70,000 MWd/tM. (author)

  13. Non destructive assay of nuclear LEU spent fuels for burnup credit application

    International Nuclear Information System (INIS)

    Lebrun, A.; Bignan, G.

    2001-01-01

    Criticality safety analysis devoted to spent fuel storage and transportation has to be conservative in order to be sure no accident will ever happen. In the spent fuel storage field, the assumption of freshness has been used to achieve the conservative aspect of criticality safety procedures. Nevertheless, after being irradiated in a reactor core, the fuel elements have obviously lost part of their original reactivity. The concept of taking into account this reactivity loss in criticality safety analysis is known as Burnup credit. To be used, Burnup credit involves obtaining evidence of the reactivity loss with a Burnup measurement. Many non destructive assays (NDA) based on neutron as well as on gamma ray emissions are devoted to spent fuel characterization. Heavy nuclei that compose the fuels are modified during irradiation and cooling. Some of them emit neutrons spontaneously and the link to Burnup is a power link. As a result, burn-up determination with passive neutron measurement is extremely accurate. Some gamma emitters also have interesting properties in order to characterize spent fuels but the convenience of the gamma spectrometric methods is very dependent on characteristics of spent fuel. In addition, contrary to the neutron emission, the gamma signal is mostly representative of the peripheral rods of the fuels. Two devices based on neutron methods but combining different NDA methods which have been studied in the past are described in detail: 1. The PYTHON device is a combination of a passive neutron measurement, a collimated total gamma measurement, and an online depletion code. This device, which has been used in several Nuclear Power Plants in western Europe, gives the average Burnup within a 5% uncertainty and also the extremity Burnup, 2. The NAJA device is an automatic device that involves three nuclear methods and an online depletion code. It is designed to cover the whole fuel assembly panel (Active Neutron Interrogation, Passive Neutron

  14. Burn up determination of IEAR-1 fuel elements by non destructive gamma ray spectrometry method

    International Nuclear Information System (INIS)

    Soares, A.J.

    1977-01-01

    Measurement of nuclear fuel burn up by non destructive gamma ray spectrometry is discussed, and results of such measurements, made at the Instituto de Energia Atomica (IEA), are given. Specifically, the burn up of an MTR (Material Testing Reactor) fuel element removed from the IEAR-1 swimming pool reactor in 1958 is evaluated from the measured Cs-137 activity, which gives a single 661,6 keV gamma ray. Due to the long decay time of the test element, no other fission decay product activity could be detected. Analysis of measurements, made with a 3'' x 3'' NaI(Tl) detector at 330 distinct points of the element, showed the total burn up to 3.3 +- -+ 0.8 mg. This is in agreement with a calculated value. As the maximum temperature of IEAR-1 fuel elements is of the order of 40 0 C, migration effects of Cs-137 was not considered, this being significant only at fuel temperature in excess of 1000 0 C [pt

  15. Validation of MOX fuel through recent BELGONUCLEAIRE international programmes

    International Nuclear Information System (INIS)

    Basselier, J.; Maldague, T.; Lippens, M.

    1997-01-01

    The paper reviews the present experience of BELGONUCLEAIRE in promoting and managing international programmes dedicated to improvement and updating of MOX fuel data bases on what concerns core physics and rod behaviour with a view of assist all MOX fuel designers and users in their validation and modelization work. All these programmes were completed or will be completed with the support of numerous international organizations deeply concerned by MOX recycling strategies. (author). 9 figs, 2 tabs

  16. Mox fuel utilization in ATR

    OpenAIRE

    下村 和生; 川太 徳夫

    1987-01-01

    ATR, a heavy-water moderated boiling-light-water cooled reactor developed in Japan, is a unique reactor with out-standing flexibility regarding nuclear fuel utilization, because it has superior properties concerning the utilization of plutonium, recovered uranium and depleted uranium. The development of this type of reactor is expected to contribute both to the stable supply of energy and to the establishment of plutonium utilization in Japan. Much effort has been and will be made on the deve...

  17. Thermal conductivity of heterogeneous LWR MOX fuels

    Science.gov (United States)

    Staicu, D.; Barker, M.

    2013-11-01

    It is generally observed that the thermal conductivity of LWR MOX fuel is lower than that of pure UO2. For MOX, the degradation is usually only interpreted as an effect of the substitution of U atoms by Pu. This hypothesis is however in contradiction with the observations of Duriez and Philiponneau showing that the thermal conductivity of MOX is independent of the Pu content in the ranges 3-15 and 15-30 wt.% PuO2 respectively. Attributing this degradation to Pu only implies that stoichiometric heterogeneous MOX can be obtained, while we show that any heterogeneity in the plutonium distribution in the sample introduces a variation in the local stoichiometry which in turn has a strong impact on the thermal conductivity. A model quantifying this effect is obtained and a new set of experimental results for homogeneous and heterogeneous MOX fuels is presented and used to validate the proposed model. In irradiated fuels, this effect is predicted to disappear early during irradiation. The 3, 6 and 10 wt.% Pu samples have a similar thermal conductivity. Comparison of the results for this homogeneous microstructure with MIMAS (heterogeneous) fuel of the same composition showed no difference for the Pu contents of 3, 5.9, 6, 7.87 and 10 wt.%. A small increase of the thermal conductivity was obtained for 15 wt.% Pu. This increase is of about 6% when compared to the average of the values obtained for 3, 6 and 10 wt.% Pu. For comparison purposes, Duriez also measured the thermal conductivity of FBR MOX with 21.4 wt.% Pu with O/M = 1.982 and a density close to 95% TD and found a value in good agreement with the estimation obtained using the formula of Philipponneau [8] for FBR MOX, and significantly lower than his results corresponding to the range 3-15 wt.% Pu. This difference in thermal conductivity is of about 20%, i.e. higher than the measurement uncertainties.Thus, a significant difference was observed between FBR and PWR MOX fuels, but was not explained. This difference

  18. Determination of burnup for IEAR-1 fuel elements by non destructive method of gamma spectrometry

    International Nuclear Information System (INIS)

    Madi Filho, T.; Holland, L.

    1982-01-01

    Burnup determination, by non-destructive gamma spectrometry of spent fuel with high and low activity of IEAR-1 reactor, using Cs-137 as burnup monitor, were done. To measure the Cs-137 distribution in these elements a Ge(Li) detector, with volume equal to 73,7 cm 3 , in two measurement systems with defined geometry and good colimation, was used. The IEA-14 taken from the core about 20 years ago, presents a gamma spectra due to Cs-137. The IEA-80, with cooling time approximately to 5 years, shows a more complex gamma spectrum due to other fission products still found in significant quantities. The IEA-14 measures were done in a measurement system used outside the reactor pool (S.I.), being the global efficiency of this system obtained by using a plane, calibrated and extense Ag-110 m source. Detailed measures of gamma transmission, using Cs-137 as a calibrated and punctiforme source, showed the high homogenity of the fuel plates. (E.G.) [pt

  19. 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)

  20. Burn of actinides in MOX fuel cells

    International Nuclear Information System (INIS)

    Martinez C, E.; Ramirez S, J. R.; Alonso V, G.

    2017-09-01

    The spent fuel from nuclear reactors is stored temporarily in dry repositories in many countries of the world. However, the main problem of spent fuel, which is its high radio-toxicity in the long term, is not solved. A new strategy is required to close the nuclear fuel cycle and for the sustain ability of nuclear power generation, this strategy could be the recycling of plutonium to obtain more energy and recycle the actinides generated during the irradiation of the fuel to transmute them in less radioactive radionuclides. In this work we evaluate the quantities of actinides generated in different fuels and the quantities of actinides that are generated after their recycling in a thermal reactor. First, we make a reference calculation with a regular enriched uranium fuel, and then is changed to a MOX fuel, varying the plutonium concentrations and determining the quantities of actinides generated. Finally, different amounts of actinides are introduced into a new fuel and the amount of actinides generated at the end of the fuel burn is calculated, in order to determine the reduction of minor actinides obtained. The results show that if the concentration of plutonium in the fuel is high, then the production of minor actinides is also high. The calculations were made using the cell code CASMO-4 and the results obtained are shown in section 6 of this work. (Author)

  1. Non-destructive assay of leached hulls in a nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Hofstetter, K.J.; Henderson, B.C.; Gray, J.H.; Huff, G.A.

    1978-01-01

    The hull monitor at the Barnwell Nuclear Fuels Plant (BNFP) will be a remotely controlled, fully automated system designed to quantitatively assay leached hulls for undissolved U and Pu. The hull monitor will assay the hulls from one metric ton of fuel per dissolver basket with the design goal of detecting 0.1% undissolved fuel and yet remain within the framework of the BNFP materials flow of five hull baskets per day. The non-destructive assay will be accomplished using a computer-based gamma-ray pulse height analysis system employing a 5 x 5 inch NaI(Tl) scintillation detector. The intense radiations from the fission product isotopes and the activation product isotopes produced in the reactor prevent direct assay of the undissolved fuel left in the hulls. The measurement will be made indirectly by demonstrating a correlation between the amount of 144 Ce undissolved and the remaining U. The isotope 144 Ce is a direct fission product with high cumulative yield. The daughter isotope 144 Pr has a gamma ray at 2.18 MeV well above other predominant radiations in the spectrum from the major interferences 60 Co, 58 Co, 95 Zr( 95 Nb), 137 Cs and 106 Ru( 106 Rh). Segmented scanning operation of the hull monitor is accomplished by rotation and vertical transversal of the hulls container past the detector station. Proper collimation and absorbers are required to maximize the 144 Ce( 144 Pr) to background ratio. A basket indexer is provided which monitors the scanning rate and ensures repositioning. The leached hull monitor system will be interfaced to a computer-based multichannel analyzer for ease of operation and data handling. A calibration basket has been fabricated to accomodate radioactive sources and inactive Zircaloy hulls

  2. Dissolution behavior of PFBR MOX fuel in nitric acid

    International Nuclear Information System (INIS)

    Kelkar, Anoop; Kapoor, Y.S.; Singh, Mamta; Meena, D.L.; Pandey, Ashish; Bhatt, R.B.; Behere, P.G.

    2017-01-01

    Present paper describes the dissolution characteristics of PFBR MOX fuel (U,Pu)O 2 in nitric acid. An overview of batch dissolution experiments, studying the percentage dissolution of uranium and plutonium in (U, Pu)O 2 MOX sintered pellets with different percentage of PuO 2 with reference to time and nitric acid concentration are described. 90% of uranium and plutonium of PFBR MOX gets dissolves in 2 hrs and amount of residue increases with the decrease in nitric acid concentration. Overall variation in percentage residue in PFBR MOX fuel after dissolution test also described. (author)

  3. 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.

  4. MOX fuel fabrication, in reactor performance and improvement

    International Nuclear Information System (INIS)

    Vliet, J. van; Deramaix, P.; Nigon, J.L.; Fournier, W.

    1998-01-01

    In Europe, MOX fuel for light water reactors (LWRs) has first been manufactured in Belgium and Germany. Belgonucleaire (BN) loaded the first MOX assembly in the BR3 Pressurised Water Reactor (PWR) in 1963. In June 1998, more than 750 tHM LWR MOX fuel assemblies were manufactured on a industrial scale in Europe without any particular difficulty relating to fuel fabrication, reactor operation or fuel behaviour. So, today plutonium recycling through MOX fuel is a mature industry, with successful operational experience and large-scale fabrication plants. In this field, COGEMA and BELGONUCLEAIRE are the main actors by operating simultaneously three complete multidesign fuel production plants: MELOX plant (in Marcoule), CADARACHE plant and P0 plant (in Dessel, Belgium). Present MOX production capacity available to COGEMA and BN fits 175 tHM per year and is to be extended to reach about 325 tHM in the year 2000. This will represent 75% of the total MOX fabrication capacity in Europe. The industrial mastery and the high production level in MOX fabrication assured by high technology processes confer to these companies a large expertise for Pu recycling. This allows COGEMA and BN to be major actors in Pu-based fuels in the coming second nuclear era with advanced fuel cycles. (author)

  5. Developments in MOX fuel pellet fabrication technology: Indian experience

    International Nuclear Information System (INIS)

    Kamath, H.S.; Majumdar, S.; Purusthotham, D.S.C.

    1998-01-01

    India is interested in mixed oxide (MOX) fuel technology for better utilisation of its nuclear fuel resources. In view of this, a programme involving MOX fuel design, fabrication and irradiation in research and power reactors has been taken up. A number of experimental irradiations in research reactors have been carried out and a few MOX assemblies of ''All Pu'' type have been loaded in our commercial BWRs at Tarapur. An island type of MOX fuel design is under study for use in PHWRs which can increase the burn-up of the fuel by more than 30% compared to natural UO 2 fuel. The MOX fuel pellet fabrication technology for the above purpose and R and D efforts in progress for achieving better fuel performance are described in the paper. The standard MOX fuel fabrication route involves mechanical mixing and milling of UO 2 and PuO 2 powders. After detailed investigations with several types of mixing and milling equipments, dry attritor milling has been found to be the most suitable for this operation. Neutron Coincident Counting (NCC) technique was found to be the most convenient and appropriate technique for quick analysis of Pu content in milled MOX powder and to know Pu mixing is homogenous or not. Both mechanical and hydraulic presses have been used for powder compaction for green pellet production although the latter has been preferred for better reproducibility. Low residue admixed lubricants have been used to facilitate easy compaction. The normal sintering temperature used in Nitrogen-Hydrogen atmosphere is between 1600 deg. C to 1700 deg. C. Low temperature sintering (LTS) using oxidative atmospheres such as carbon dioxide, Nitrogen and coarse vacuum have also been investigated on UO 2 and MOX on experimental scale and irradiation behaviour of such MOX pellets is under study. Ceramic fibre lined batch furnaces have been found to be the most suitable for MOX pellet production as they offer very good flexibility in sintering cycle, and ease of maintainability

  6. MOX fuel development: Experience in Argentina

    International Nuclear Information System (INIS)

    Marchi, D.E.; Adelfang, P.; Menghini, J.E.

    1999-01-01

    Since 1973, when a laboratory conceived for the safe manipulation of a few hundred grams of plutonium was built, the CNEA (Argentinean Atomic Energy Commission) has been involved in the small-scale development of MOX fuel technology. The plutonium laboratory consists in a glove box facility (α Facility) featuring the necessary equipment to prepare MOX fuel rods for experimental irradiations and to carry out studies on preparative processes development and chemical and physical characterization. The irradiation of the first prototypes of (U,Pu)O 2 fuels fabricated in Argentina began in 1986. These experiments were carried out in the HFR (High Flux Reactor)- Petten , Holland. The rods were prepared and controlled in the CNEA's a Facility. The post-irradiation examinations (PIE) were performed in the KFK (Kernforschungszentrum Karlsruhe), Germany and the JRC (Joint Research Center), Petten. In the period 1991-1995, the development of new laboratory methods of co-conversion of uranium and plutonium were carried out: reverse strike co-precipitation of ADU-Pu(OH) 4 and direct denitration using microwaves. The reverse strike process produced pellets with a high sintered density, excellent micro-homogeneity and good solubility in nitric acid. Liquid wastes showed a very low content of actinides and the process is easy to operate in a glove box environment. The microwave direct denitration was optimized with uranium alone and the conditions to obtain high density pellets, with a good microstructure, without using a milling step, have been developed. At present, new experiments are being carried out to improve the reverse strike co-precipitation process and direct microwave denitration. A new glove box is being installed at the plutonium laboratory, this glove box has process equipment designed to recover scrap from previous fabrication campaigns, and to co-convert mixed U-Pu solutions by direct microwave denitration. (author)

  7. Power ramp tests of BWR-MOX fuels

    International Nuclear Information System (INIS)

    Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

    1996-01-01

    Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

  8. Preliminary nuclear design for test MOX Fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

    1997-10-01

    As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

  9. Non-destructive test for VHTR fuel using 160kV X-ray system in Hotcell

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jun; Yoo, Boung Ok; Choo, Yong sun; Baik Sang youl; Kim, Hee Moon; Ahn, Sang Bok [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The research for VHTR which is one of the next generation reactor has been actively carried out. As a part of the research for VHTR, an irradiation examination for the VHTR fuel was performed to confirm an in-pile behavior in HANARO. The non-destructive test for the irradiated fuel is very important to understand the in-pile behavior of the fuel. Especially, the X-ray system is useful to observe the fuel shape without destruction. A dimensional change and defect of the fuel can be confirmed thorough the Xray system. Also, using the 3-D software and CT technology, the fuel shape can be intuitionally observed. The 450kV and 160kV X-ray system were installed and operated in IMEF hotcell. The 160kV X-ray system relatively using a low voltage is suitable to a small scale sample. And high resolution images can be obtained. In this study, the non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. Through these test, the possibility for the X-ray inspection of irradiated fuel was confirmed. The non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. The clear images of the irradiated coated particle were produced without the radiation damage during the Xray inspection. The X-ray images of the VHTR fuel will be utilized as the in-pile performance validation data.

  10. Non-destructive test for VHTR fuel using 160kV X-ray system in Hotcell

    International Nuclear Information System (INIS)

    Kim, Young Jun; Yoo, Boung Ok; Choo, Yong sun; Baik Sang youl; Kim, Hee Moon; Ahn, Sang Bok

    2016-01-01

    The research for VHTR which is one of the next generation reactor has been actively carried out. As a part of the research for VHTR, an irradiation examination for the VHTR fuel was performed to confirm an in-pile behavior in HANARO. The non-destructive test for the irradiated fuel is very important to understand the in-pile behavior of the fuel. Especially, the X-ray system is useful to observe the fuel shape without destruction. A dimensional change and defect of the fuel can be confirmed thorough the Xray system. Also, using the 3-D software and CT technology, the fuel shape can be intuitionally observed. The 450kV and 160kV X-ray system were installed and operated in IMEF hotcell. The 160kV X-ray system relatively using a low voltage is suitable to a small scale sample. And high resolution images can be obtained. In this study, the non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. Through these test, the possibility for the X-ray inspection of irradiated fuel was confirmed. The non-destructive test for the unirradiated and irradiated VHTR fuel were performed using the 160kV X-ray system. The clear images of the irradiated coated particle were produced without the radiation damage during the Xray inspection. The X-ray images of the VHTR fuel will be utilized as the in-pile performance validation data.

  11. Characterisation of nuclear dispersion fuels. The non-destructive examination of silicon carbide by selenium immersion

    Energy Technology Data Exchange (ETDEWEB)

    Ambler, J.F.R.; Ferguson, I.F.

    1974-07-15

    The non-destructive microscopic examination of silicon-carbide-coated spheres containing uranium carbide, which involves immersing the coated spheres in selenium, is particularly suited for the examination of flaws in the coats but it is not possible to measure coating thicknesses by this method. Some coats are found to be opaque and this is related to their porosity. (auth)

  12. Transport of MOX fuel from Europe to Japan; Transport de combustible mox d' Europe vers le Japon

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    The MOX fuel transports from Europe to Japan represent a main part in the implementing of the Japan nuclear program. They complement the 160 transports of spent fuels realized from Japan to Europe and the vitrified residues return from France to Japan. In this framework the document presents the MOX fuel, the use of the MOX fuel in reactor, the proliferation risks, the MOX fuel transport to Japan, the public health, the transport regulations, the safety and the civil liability. (A.L.B.)

  13. Transport of MOX fuel from Europe to Japan; Transport de combustible mox d' Europe vers le Japon

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    The MOX fuel transports from Europe to Japan represent a main part in the implementing of the Japan nuclear program. They complement the 160 transports of spent fuels realized from Japan to Europe and the vitrified residues return from France to Japan. In this framework the document presents the MOX fuel, the use of the MOX fuel in reactor, the proliferation risks, the MOX fuel transport to Japan, the public health, the transport regulations, the safety and the civil liability. (A.L.B.)

  14. Foundations for the definition of MOX fuel quality requirements

    International Nuclear Information System (INIS)

    Bairiot, H.; Deramaix, P.; Vanderborck, Y.

    1991-01-01

    The quality of uranium-plutonium mixed oxide (MOX) fuel, as of any nuclear fuel, depends on the design optimization and on the fabrication process stability. The design optimization is essentially based on feed-back from irradiation experience through engineering assessment of the results; the stability of the process is necessary to justify minimal uncertainty margins in the fuel design. Since MOX fuel is quite similar to UO 2 fuel, the lessons learned from UO 2 fuels can complement the MOX experimental data base. MOX is however different from UO 2 fuel in some respects, among others: the industrial fabrication scale is a factor 10 lower than for UO 2 fuel, the fuel enrichment process takes place in the manufacturing plant, the radioactivity of Pu imposes handling constraints, Pu ages quite rapidly, altering its isotopic composition during storage, the incorporation of Pu alters the material physics and neutronic characteristics of the fuel. In this perspective, the paper outlines some quality attributes for which MOX fuel may or even must depart form UO 2 fuel. (orig.)

  15. Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

    Energy Technology Data Exchange (ETDEWEB)

    Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V

    2000-07-01

    One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX {yields} MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

  16. Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

    International Nuclear Information System (INIS)

    Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V.

    2000-01-01

    One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOXMOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

  17. MOX fuel fabrication and utilisation in LWRs worldwide

    International Nuclear Information System (INIS)

    Provost, J.-L.; Schrader, M.; Nomura, S.

    2000-01-01

    Early in the development of the nuclear programme, a large part of the countries using nuclear energy has studied the reprocessing and recycling option in order to develop a safe conditioning of fission products and to recycle fissile materials in reactors. In the sixties, the feasibility of recycling plutonium in LWRs has been successfully demonstrated by several experimentations of MOX rod irradiations in different countries. Based on the background of the MOX behaviour collected during the seventies and on the results of the important MOX experimentation program implemented during this period, a large part of the European utilities decided at the beginning of the eighties to use MOX fuel in LWRs on an industrial scale. The main goals of the utilities were to use as a fuel an available fissile material and to control the stockpile of separated plutonium. Today, the understanding of the behaviour of plutonium fuel has grown significantly since the launch of the first R and D programmes on LWR and FR MOX fuels. Plutonium oxide physical and neutron behaviour is well known, its modelling is now available as well as experimentally validated. Up to now, more than 750 tHM MOX fuel (more than 2000 FAs) have been loaded in 29 PWRs and in 2 BWRs in Europe, corresponding to the recycling of about 35 t of plutonium. Reprocessing/recycling technology has reached maturity in the main nuclear industry countries. Spent fuel reprocessing and recycling of the separated fissile materials remains the main option for the back-end cycle. Today, the operation of MOX-recycling LWRs is considered satisfactory. Experience feedback shows that, in global terms, MOX cores behaviour is equivalent to that of UO 2 cores in terms of operation and safety. (author)

  18. Nuclear fuel and/or fertile material element suitable for non-destructive determination of burn-up

    International Nuclear Information System (INIS)

    Muench, E.

    1976-01-01

    The invention refers to a nuclear fuel and/or fertile material element suitable for non-destructive burn-up analysis, where an isotope or a mixture of isotopes capable of being activated is provided for measuring the intensity of radiation emitted from radioactive nuclides, especially the intensity of gamma rays. The half-life of radioactive decay of the isotope or the mixture mentioned above after being activated is sufficiently large compared with the irradiation of the fuel and/or fertile material element in the nuclear reactor. (orig.) [de

  19. Power ramp tests of MOX fuel rods. HBWR irradiation with the instrument rig, IFA-591

    International Nuclear Information System (INIS)

    Ozawa, Takayuki; Abe, Tomoyuki

    2006-03-01

    Plutonium-uranium mixed oxide (MOX) fuel rods of instrumental rig IFA-591 were ramped in HBWR to study the Advanced Thermal Reactor (ATR) MOX fuel behavior during transient operation and to determine a failure threshold of the MOX fuel rods. Eleven segments were base-irradiated in ATR 'FUGEN' up to 18.4 GWd/t. Zirconium liner claddings were adopted for four segments of them. As the results of non-destructive post irradiation examinations (PIEs) after the base-irradiation and before the ramp tests, no remarkable behavior affecting the integrity of fuel assembly and fuel rod was confirmed. All segments to be used for the ramp tests, which consisted of the multi-step ramp tests and the single-step ramp tests, had instrumentations for in-pile measurements of cladding elongation or plenum pressure, and heated up to the maximum linear power of 58.3-68.4 kW/m without failure. The major results of ramp tests are as follows: There is no difference in PCMI behaviors between two type rods of Zry-2 and Zirconium liner claddings from the in-pile measurements of cladding elongation and plenum pressure. The computations of cladding elongation and inner pressure gave slightly lower elongation and pressure than the in-pile measurements during the ramp-test. However, the cladding relaxation during the power hold was in good agreement, and the fission gas release behavior during cooling down could be evaluated by taking into account the relaxation of contact pressure between pellet and cladding. Although the final power during IFA-591 ramp tests reached the higher linear power than the failure threshold power of UO 2 fuel rods, no indication of fuel failure was observed during the ramp tests. The cladding relaxation due to the creep deformation of the MOX pellets at high temperature could be confirmed at the power steps during the multi-ramp test. The fission gas release due to the emancipation from PCMI stress was observed during the power decreasing. The burn-up dependence could be

  20. Study on high performance MOX fuel and core design in full MOX ABWR(1) by GNF-J

    International Nuclear Information System (INIS)

    Izutsu, Sadayuki; Goto, Daisuke; Saeki, Jun; Kokubun, Takehiro; Yokoya, Jun

    2003-01-01

    The concepts of high-performance MOX fuel using 10x10 lattices suitable for full-MOX ABWR are shown in this paper, in which average discharge exposure is extended up to 45 GWd/t with heavy-metal inventory increased over current MOX, reducing the number of refueling bundles, resulting in fuel cycle cost reduction and core performance satisfaction. Also, the increase of Pu inventory is taken into account from the viewpoint to extend the flexibility of MOX fuel utilization. (author)

  1. Top-MOX fuel solution: strategies, challenges, opportunities

    International Nuclear Information System (INIS)

    Breitenstein, P.; Vo Van, V.

    2014-01-01

    TOP-MOX is a nuclear fuel solution and product developed by AREVA and successfully implemented in Europe. It allows utilities burning plutonium (instead of enriched uranium) even when this plutonium is not stemming from own reprocessed used fuel - that is third party plutonium. The important challenges for utilities along with TOP-MOX implementation are legal/patrimonial Pu-ownership issues and general economical aspects. Available sponsorship of such plutonium permits UO2 competitive market prices. For new MOX customers licensing and technical aspects come along. Further AREVA proposes a flexible solution which is called 'TOP-MOX pre-cycling'. This involves making available third party plutonium for fuel fabrication and reactor use pending the utilities' final strategic fuel cycle decision. The paper gives insight into and analyses the impacts of allowing customers the implementation of a TOP-MOX program with focus on Pu-ownership, economics, technical and legal aspects as well as the impact on used MOX management and final waste management. (authors)

  2. Fuel rod design by statistical methods for MOX fuel

    International Nuclear Information System (INIS)

    Heins, L.; Landskron, H.

    2000-01-01

    Statistical methods in fuel rod design have received more and more attention during the last years. One of different possible ways to use statistical methods in fuel rod design can be described as follows: Monte Carlo calculations are performed using the fuel rod code CARO. For each run with CARO, the set of input data is modified: parameters describing the design of the fuel rod (geometrical data, density etc.) and modeling parameters are randomly selected according to their individual distributions. Power histories are varied systematically in a way that each power history of the relevant core management calculation is represented in the Monte Carlo calculations with equal frequency. The frequency distributions of the results as rod internal pressure and cladding strain which are generated by the Monte Carlo calculation are evaluated and compared with the design criteria. Up to now, this methodology has been applied to licensing calculations for PWRs and BWRs, UO 2 and MOX fuel, in 3 countries. Especially for the insertion of MOX fuel resulting in power histories with relatively high linear heat generation rates at higher burnup, the statistical methodology is an appropriate approach to demonstrate the compliance of licensing requirements. (author)

  3. Development of MOX facilities and the impact on the nuclear fuel markets

    International Nuclear Information System (INIS)

    Patterson, J.

    1990-01-01

    Mixed-oxide (MOX) fuel is nearing maturity as a fuel supply option. This paper briefly reviews the history and current status of the MOX fuel market, including the projected increase in demand for MOX fuel as more plutonium becomes available from the operation of commercial irradiated fuel reprocessing plants in Europe. The uncertainties of such projected demand are discussed, together with the anticipated requirements from the next generation of MOX fabrication plants. The impact of the growing demand for MOX fuel is assessed in the traditional sectors of the uranium fuel cycle. Finally, the author turns to a generalized treatment of the economic aspects of MOX fuel utilization, showing the financially attractive regimes of MOX use which will benefit nuclear power utilities and continue to ensure that MOX fuel can consolidate its position as a mature fuel supply option in those countries that have opted to recycle their spent fuel

  4. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    Provost, Jean-Luc; Debes, Michel

    2005-01-01

    Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

  5. Specific features of the determination of the pellet-cladding gap of the fuel rods by non-destructive method

    International Nuclear Information System (INIS)

    Amosov, S.V.; Pavlov, S.V.

    2002-01-01

    This report describes the specific features of determining the pellet-cladding gap of the irradiated WWER-1000 fuel rods by nondestructive method. The method is based on the elastic radial deformation of the cladding up to its contact with the fuel. The value of deformation of cladding till its contacting fuel when radial force changes from F max to 0 is proposed as a measuring parameter for determination of the diametrical gap. Because of the features of compression method, the obtained gap value is not analog of the gap measured on micrograph of the fuel rod cross-section. Results of metallography can provide only qualitative evaluation of its method efficiency. Comparison of the values determined by non-destructive method and metallography for WWER-1000 fuel rods with burnup from 25 to 55 MWd/kg U testified that the results of compression method can be used as a low estimate of the pellet-cladding gap value. (author)

  6. Mixed Reload Design Using MOX and UOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Ramon, Ramirez Sanchez J.; Perry, R.T.

    2002-01-01

    As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10 X 10 BWR fuel assemblies but different fissile material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO 2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fissile plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO 2 fresh fuel were tested to verify the shutdown margin, the UO 2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper. (authors)

  7. 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)

  8. Advanced high throughput MOX fuel fabrication technology and sustainable development

    International Nuclear Information System (INIS)

    Krellmann, Juergen

    2005-01-01

    The MELOX plant in the south of France together with the La Hague reprocessing plant, are part of the two industrial facilities in charge of closing the nuclear fuel cycle in France. Started up in 1995, MELOX has since accumulated a solid know-how in recycling plutonium recovered from spent uranium fuel into MOX: a fuel blend comprised of both uranium and plutonium oxides. Converting recovered Pu into a proliferation-resistant material that can readily be used to power a civil nuclear reactor, MOX fabrication offers a sustainable solution to safely take advantage of the plutonium's high energy content. Being the first large-capacity industrial facility dedicated to MOX fuel fabrication, MELOX distinguishes itself from the first generation MOX plants with high capacity (around 200 tHM versus around 40 tHM) and several unique operational features designed to improve productivity, reliability and flexibility while maintaining high safety standards. Providing an exemplary reference for high throughput MOX fabrication with 1,000 tHM produced since start-up, the unique process and technologies implemented at MELOX are currently inspiring other MOX plant construction projects (in Japan with the J-MOX plant, in the US and in Russia as part of the weapon-grade plutonium inventory reduction). Spurred by the growing international demand, MELOX has embarked upon an ambitious production development and diversification plan. Starting from an annual level of 100 tons of heavy metal (tHM), MELOX demonstrated production capacity is continuously increasing: MELOX is now aiming for a minimum of 140 tHM by the end of 2005, with the ultimate ambition of reaching the full capacity of the plant (around 200 tHM) in the near future. With regards to its activity, MELOX also remains deeply committed to sustainable development in a consolidated involvement within AREVA group. The French minister of Industry, on August 26th 2005, acknowledged the benefits of MOX fuel production at MELOX: 'In

  9. Emission computer tomography on a Dodewaard mixed oxide fuel pin. Comparative PIE work with non-destructive and destructive techniques

    Energy Technology Data Exchange (ETDEWEB)

    Buurveld, H.A.; Dassel, G.

    1993-12-01

    A nondestructive technique as well as a destructive PIE technique have been used to verify the results obtained with a newly 8-e computer tomography (GECT) system. Multi isotope Scanning (MIS), electron probe micro analysis (EPMA) and GECT were used on a mixed oxide (MOX) fuel rod from the Dodewaard reactor with an average burnup of 24 MWd/kg fuel. GECT shows migration of Cs to the periphery of fuel pellets and to radial cracks and pores in the fuel, whereas MIS shows Cs migration to pellet interfaces. The EPMA technique appeared not to be useful to show migration of Cs but, it shows the distribution of fission products from Pu. EPMA clearly shows the distribution of fission products from Pu, but did not reveal the Cs-migration. (orig./HP)

  10. Studies of Flexible MOX/LEU Fuel Cycles

    International Nuclear Information System (INIS)

    Adams, M.L.; Alonso-Vargas, G.

    1999-01-01

    This project was a collaborative effort involving researchers from Oak Ridge National Laboratory and North Carolina State University as well as Texas A and M University. The background, briefly, is that the US is planning to use some of its excess weapons Plutonium (Pu) to make mixed-oxide (MOX) fuel for existing light-water reactors (LWRs). Considerable effort has already gone into designing fuel assemblies and core loading patterns for the transition from full-uranium cores to partial-MOX and full-MOX cores. However, these designs have assumed that any time a reactor needs MOX assemblies, these assemblies will be supplied. In reality there are many possible scenarios under which this supply could be disrupted. It therefore seems prudent to verify that a reactor-based Pu-disposition program could tolerate such interruptions in an acceptable manner. Such verification was the overall aim of this project. The task assigned to the Texas A and M team was to use the HELIOS code to develop libraries of two-group homogenized cross sections for the various assembly designs that might be used in a Westinghouse Pressurized Water Reactor (PWR) that is burning weapons-grade MOX fuel. The NCSU team used these cross sections to develop optimized loading patterns under several assumed scenarios. Their results are documented in a companion report

  11. Code Analyses Supporting PIE of Weapons-Grade MOX Fuel

    International Nuclear Information System (INIS)

    Ott, Larry J.; Bevard, Bruce Balkcom; Spellman, Donald J.; McCoy, Kevin

    2010-01-01

    The U.S. Department of energy has decided to dispose of a portion of the nation's surplus weapons-grade plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating the fuel in commercial power reactors. Four lead test assemblies (LTAs) were manufactured with weapons-grade mixed oxide (WG-MOX) fuel and irradiated in the Catawba Nuclear Station Unit 1, to a maximum fuel rod burnup of ∼47.3 GWd/MTHM. As part of the fuel qualification process, five rods with varying burnups and initial plutonium contents were selected from one assembly and shipped to the Oak Ridge National Laboratory (ORNL) for hot cell examination. ORNL has provided analytical support for the post-irradiation examination (PIE) of these rods via extensive fuel performance modeling which has aided in instrument settings and PIE data interpretation. The results of these fuel performance simulations are compared in this paper with available PIE data.

  12. Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Miyanishi, Hideyuki; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi.

    1991-11-01

    Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85 Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/V f or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO 2 fuel, C (counts/h) the radioactivity of 85 Kr at plenum of the tested fuel rod and V f (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

  13. Non-destructive evaluation of the cladding thickness in LEU fuel plates by accurate ultrasonic scanning technique

    Energy Technology Data Exchange (ETDEWEB)

    Borring, J.; Gundtoft, H.E.; Borum, K.K.; Toft, P. [Riso National Lab. (Denmark)

    1997-08-01

    In an effort to improve their ultrasonic scanning technique for accurate determination of the cladding thickness in LEU fuel plates, new equipment and modifications to the existing hardware and software have been tested and evaluated. The authors are now able to measure an aluminium thickness down to 0.25 mm instead of the previous 0.35 mm. Furthermore, they have shown how the measuring sensitivity can be improved from 0.03 mm to 0.01 mm. It has now become possible to check their standard fuel plates for DR3 against the minimum cladding thickness requirements non-destructively. Such measurements open the possibility for the acceptance of a thinner nominal cladding than normally used today.

  14. Non-destructive evaluation of the cladding thickness in LEU fuel plates by accurate ultrasonic scanning technique

    International Nuclear Information System (INIS)

    Borring, J.; Gundtoft, H.E.; Borum, K.K.; Toft, P.

    1997-01-01

    In an effort to improve their ultrasonic scanning technique for accurate determination of the cladding thickness in LEU fuel plates, new equipment and modifications to the existing hardware and software have been tested and evaluated. The authors are now able to measure an aluminium thickness down to 0.25 mm instead of the previous 0.35 mm. Furthermore, they have shown how the measuring sensitivity can be improved from 0.03 mm to 0.01 mm. It has now become possible to check their standard fuel plates for DR3 against the minimum cladding thickness requirements non-destructively. Such measurements open the possibility for the acceptance of a thinner nominal cladding than normally used today

  15. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    Nagano, M.; Sakurai, S.; Yamaguchi, H.

    1997-01-01

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  16. Micro-Reactor Physics of MOX-Fueled Core

    International Nuclear Information System (INIS)

    Takeda, T.

    2001-01-01

    Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design

  17. Progress in researches on MOX fuel pellet producing technology in China

    International Nuclear Information System (INIS)

    Hu Xiaodan

    2010-01-01

    Being the key section of nuclear-fuel cycle, the producing technology of MOX(UO 2 -PuO 2 ) fuel had driven to maturity in France, England, Russia, Belgium, etc. MOX fuel had been applied in FBR and LWR successfully in those countries. With the rapidly developing of nuclear-generated power, the MOX fuel for FBR and LWR was active demanded in China. However, the producing technology of MOX fuel developed slowly. During the period of 'the seventh five year's project', MOX fuel pellet was produced by mechanically mixed method and oxalate deposited method, respectively. Parts of cool performance of MOX fuel pellet produced by oxalate deposited method reached the qualification of fuel for FBR. During the period of 'the ninth five year's project' and 'the tenth five year's project', the technical route of producing MOX fuel was determined, and the test line of producing MOX fuel was built preliminarily. In the same time, the producing technology and analyzing technology of MOX fuel pellet by mechanically mixed was studied roundly, and the representative analogue pellet(UO 2 -CeO 2 ) was produced. That settled the supporting technology for the commercial process and research of MOX fuel rod and MOX fuel module. (authors)

  18. Overview of safeguards aspects related to MOX fuel

    International Nuclear Information System (INIS)

    Heinonen, O.J.; Murakami, K.; Shea, T.

    2000-01-01

    Recent developments in the light of the IAEA verification requirements for MOX fuel at reactors and bulk handling facilities are discussed. Impact of the Additional Protocol and Integrated Safeguards System is briefly addressed. Agency's work undertaken with regard to the nuclear arms control and reduction is presented. (author)

  19. Optimization of MOX fuel cycles in pebble bed HTGR

    International Nuclear Information System (INIS)

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

    Compared with light water reactor (LWR), the pebble bed high temperature gas-cooled reactor (HTGR) is able to operate in a full mixed oxide (MOX) fuelled core without significant change to core structure design. Based on a reference design of 250 MW pebble bed HTGR, four MOX fuel cycles were designed and evaluated by VSOP program package, including the mixed Pu-U fuel pebbles and mixed loading of separate Pu-pebbles and U-pebbles. Some important physics features were investigated and compared for these four cycles, such as the effective multiplication factor of initial core, the pebble residence time, discharge burnup, and temperature coefficients. Preliminary results show that the overall performance of one case is superior to other equivalent MOX fuel cycles on condition that uranium fuel elements and plutonium fuel elements are separated as the different fuel pebbles and that the uranium fuel elements are irradiated longer in the core than the plutonium fuel elements, and the average discharge burnup of this case is also higher than others. (authors)

  20. Review of research on non-destructive analysis of reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Ruby, S L; Gardner, D G [Westinghouse Electric Corp. (United States). Atomic Power Dept.

    1959-01-01

    This report contains a selected bibliography on the nondestructive assay of irradiated and unirradiated reactor fuel assemblies. The report also includes a discussion and evaluation of these papers in the light of the problems presented by variations in enrichment, alloying, cladding, geometrical arrangement, etc., among the many kinds of fuel assemblies. Suggestions for further development work are made. (author).

  1. Calculated and experimental substantiation of the thermal method for non-destructive testing of fuel elements

    International Nuclear Information System (INIS)

    Maksimov, N.M.; Soldatenko, V.A.; Petrovichev, V.I.; Salimov, S.E.; Aleksandrov, K.A.; Kurov, D.A.

    1985-01-01

    The main systems and methods of thermal testing, their potentialities and advantages, thermal irradiation photodetectors are described. Possible fields of application of thermal testing in nuclear engineering are discussed. Calculations of the fuel element nonstationary temperature field in the three-dimensional geometry in the presence of such an effect as fuel exfaliation from cladding are presented. The developed method and equipment for fuel element thermal testing are described. Preliminary experimental data being in agreement with the calculated ones and opening the prospects for flaw detecting are presened

  2. Non-destructive analysis of spent nuclear fuel; Nedestruktivno odredjivanje istrosenog nuklearnog goriva

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D [Institute of Nuclear Sciences Boris Kidric, Fizicka laboratorija, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Nondestructive analysis of fuel elements dealt with determining the isotope contents which provide information about the burnup level, quantities of fission products and neutron-multiplication properties of the irradiated fuel. Methods for determination of the isotope ratio of the spent fuel are both numerical and experimental. This report deals with the experimental method. This means development of the experimental methods for direct measurement of the isotope content. A number of procedures are described: measurements of {alpha}, {beta} and {gamma} activities of the isotopes; measurement of secondary effects of nuclear reactions with thermal neutrons and fast neutrons; measurement of cross sections; detection of prompt and delayed neutrons.

  3. From Russian weapons grade plutonium to MOX fuel

    International Nuclear Information System (INIS)

    Braehler, G.; Kudriavtsev, E.G.; Seyve, C.

    1997-01-01

    The April 1996, G7 Moscow Summit on nuclear matters provided a political framework for one of the most current significant challenges: ensuring a consistent answer to the weapons grade fissile material disposition issue resulting from the disarmament effort engaged by both the USA and Russia. International technical assessments have showed that the transformation of Weapons grade Plutonium in MOX fuel is a very efficient, safe, non proliferant and economically effective solution. In this regard, COGEMA and SIEMENS, have set up a consistent technical program properly addressing incineration of weapons grade plutonium in MOX fuels. The leading point of this program would be the construction of a Weapons grade Plutonium dedicated MOX fabrication plant in Russia. Such a plant would be based on the COGEMA-SIEMENS industrial capabilities and experience. This facility would be operated by MINATOM which is the partner for COGEMA-SIEMENS. MINATOM is in charge of coordination of the activity of the Russian research and construction institutes. The project take in account international standards for non-proliferation, safety and waste management. France and Germany officials reasserted this position during their last bilateral summits held in Fribourg in February and in Dijon in June 1996. MINATOM and the whole Russian nuclear community have already expressed their interest to cooperate with COGEMA-SIEMENS in the MOX field. This follows governmental-level agreements signed in 1992 by French, German and Russian officials. For years, Russia has been dealing with research and development on MOX fabrication and utilization. So, the COGEMA-SIEMENS MOX proposal gives a realistic answer to the management of weapons grade plutonium with regard to the technical, industrial, cost and schedule factors. (author)

  4. Achieving High Burnup Targets With Mox Fuels: Techno Economic Implications

    International Nuclear Information System (INIS)

    Clement Ravi Chandar, S.; Sivayya, D.N.; Puthiyavinayagam, P.; Chellapandi, P.

    2013-01-01

    For a typical MOX fuelled SFR of power reactor size, Implications due to higher burnup have been quantified. Advantages: – Improvement in the economy is seen upto 200 GWd/ t; Disadvantages: – Design changes > 150 GWd/ t bu; – Need for 8/ 16 more fuel SA at 150/ 200 GWd/ t bu; – Higher enrichment of B 4 C in CSR/ DSR at higher bu; – Reduction in LHR may be required at higher bu; – Structural material changes beyond 150 GWd/ t bu; – Reprocessing point of view-Sp Activity & Decay heat increase. Need for R & D is a must before increasing burnup. bu- refers burnup. Efforts to increase MOX fuel burnup beyond 200 GWd/ t may not be highly lucrative; • MOX fuelled FBR would be restricted to two or four further reactors; • Imported MOX fuelled FBRs may be considered; • India looks towards launching metal fuel FBRs in the future. – Due to high Breeding Ratio; – High burnup capability

  5. EURATOM safeguards efforts in the development of spent fuel verification methods by non-destructive assay

    Energy Technology Data Exchange (ETDEWEB)

    Matloch, L.; Vaccaro, S.; Couland, M.; De Baere, P.; Schwalbach, P. [Euratom, Communaute europeenne de l' energie atomique - CEEA (European Commission (EC))

    2015-07-01

    The back end of the nuclear fuel cycle continues to develop. The European Commission, particularly the Nuclear Safeguards Directorate of the Directorate General for Energy, implements Euratom safeguards and needs to adapt to this situation. The verification methods for spent nuclear fuel, which EURATOM inspectors can use, require continuous improvement. Whereas the Euratom on-site laboratories provide accurate verification results for fuel undergoing reprocessing, the situation is different for spent fuel which is destined for final storage. In particular, new needs arise from the increasing number of cask loadings for interim dry storage and the advanced plans for the construction of encapsulation plants and geological repositories. Various scenarios present verification challenges. In this context, EURATOM Safeguards, often in cooperation with other stakeholders, is committed to further improvement of NDA methods for spent fuel verification. In this effort EURATOM plays various roles, ranging from definition of inspection needs to direct participation in development of measurement systems, including support of research in the framework of international agreements and via the EC Support Program to the IAEA. This paper presents recent progress in selected NDA methods. These methods have been conceived to satisfy different spent fuel verification needs, ranging from attribute testing to pin-level partial defect verification. (authors)

  6. A programmatic approach for implementing MOX fuel operation in advanced and existing boiling water reactors

    International Nuclear Information System (INIS)

    Ehrlich, E.H.; Knecht, P.D.; Shirley, N.C.; Wadekamper, D.C.

    1996-01-01

    This paper describes a programmatic overview of the elements and issues associated with MOX fuel utilization. Many of the dominant considerations and integrated relationships inherent in initiating MOX fuel utilization in BWRs or the ABWR with partial or full MOX core designs are discussed. The most significant considerations in carrying out a MOX implementation program, while achieving commercially desirable fuel cycles and commercially manageable MOX fuel fabrication, testing, qualification, and licensing support activities, are described. The impact of politics and public influences and the necessary role of industry and government contributions are also discussed. (J.P.N.)

  7. Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

    International Nuclear Information System (INIS)

    Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

    2004-01-01

    As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

  8. 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.

  9. Non-Destructive Spent Fuel Characterization with Semiconducting Gallium Arsinde Neutron Imaging Arrays

    International Nuclear Information System (INIS)

    McGregor, Douglas S.; Gersch, Holly K.; Sanders, Jeffrey D.; Lee, John C.; Hammig, Mark D.; Hartman, Michael R.; Yong Hong Yang; Klann, Raymond T.; Elzen, Brian Van Der; Lindsay, John T.; Simpson, Philip A.

    2002-01-01

    High resistivity bulk grown GaAs has been used to produce thermal neutron imaging devices for use in neutron radiography and characterizing burnup in spent fuel. The basic scheme utilizes a portable Sb/Be source for monoenergetic (24 keV) neutron radiation source coupled to an Fe filter with a radiation hard B-coated pixellated GaAs detector array as the primary neutron detector. The coated neutron detectors have been tested for efficiency and radiation hardness in order to determine their fitness for the harsh environments imposed by spent fuel. Theoretical and experimental results are presented, showing detector radiation hardness, expected detection efficiency and the spatial resolution from such a scheme. A variety of advanced neutron detector designs have been explored, with experimental results achieving 13% thermal neutron detection efficiency while projecting the possibility of over 30% thermal neutron detection efficiency

  10. The role of Monte Carlo burnup calculations in quantifying plutonium mass in spent fuel assemblies with non-destructive assay

    Energy Technology Data Exchange (ETDEWEB)

    Galloway, Jack D.; Tobin, Stephen J.; Trellue, Holly R.; Fensin, Michael L. [Los Alamos National Laboratory, Los Alamos, (United States)

    2011-12-15

    The Next Generation Safeguards Initiate (NGSI) of the United States Department of Energy has funded a multi-laboratory/university collaboration to quantify plutonium content in spent fuel (SF) with non-destructive assay (NDA) techniques and quantify the capability of these NDA techniques to detect pin diversions from SF assemblies. The first Monte Carlo based spent fuel library (SFL) developed for the NGSI program contained information for 64 different types of SF assemblies (four initial enrichments, burnups, and cooling times). The maximum amount of fission products allowed to still model a 17x17 Westinghouse pressurized water reactor (PWR) fuel assembly with four regions per fuel pin was modelled. The number of fission products tracked was limited by the available memory. Studies have since indicated that additional fission product inclusion and asymmetric burning of the assembly is desired. Thus, an updated SFL has been developed using an enhanced version of MCNPX, more powerful computing resources, and the Monte Carlo-based burnup code Monteburns, which links MCNPX to a depletion code and models a representative 1 Division-Slash 8 core geometry containing one region per fuel pin in the assemblies of interest, including a majority of the fission products with available cross sections. Often in safeguards, the limiting factor in the accuracy of NDA instruments is the quality of the working standard used in calibration. In the case of SF this is anticipated to also be true, particularly for several of the neutron techniques. The fissile isotopes of interest are co-mingled with neutron absorbers that alter the measured count rate. This paper will quantify how well working standards can be generated for PWR spent fuel assemblies and also describe the spatial plutonium distribution across an assembly. More specifically we will demonstrate how Monte Carlo gamma measurement simulations and a Monte Carlo burnup code can be used to characterize the emitted gamma

  11. Non-destructive Engineering

    International Nuclear Information System (INIS)

    Ko, Jin Hyeon; Ryu, Taek In; Ko, Jun Bin; Hwang, Yong Hwa

    2006-08-01

    This book gives descriptions of non-destructive engineering on outline of non-destructive test, weld defects, radiographic inspection radiography, ultrasonic inspection, magnetic particle testing, liquid penetrant testing, eddy current inspection method, strain measurement, acoustic emission inspection method, other non-destructive testing like leakage inspection method, and non-destructive mechanics for fault analysis such as Griffiths creaking theory, and stress analysis of creaking.

  12. A non-destructive technique for qualitative and attribute verification of spent fuel elements

    International Nuclear Information System (INIS)

    Boeck, H.; Mueller, H.; Vana, N.

    1982-07-01

    Comparing the experiments using active and passive detectors with two types of spent fuel elements each method has its specific advantage or drawback. Active methods with fission or ionization chambers give immediate results in the facility and can be evaluated on the spot. The equipment is rather expensive, difficult to transport and needs experienced personnel. Problems may also arise with importing enriched uranium included in the fission chamber into various countries or with utilities concerned to insert fissile material in spent fuel pools. Passive methods using solid state track recorders and thermoluminescence dosimeters are easy to apply, the detectors are cheap, the detectors do not need a power supply, the transport of the equipment is very simple, and the measurements can be performed with less experienced personnel. However, the results are not obvious immediately, the detectors have to be evaluated in a laboratory by experienced personnel. With SSTR a permanent record can be kept in the files. The SSTR have to be exposed for a longer period being both an advantage and disadvantage. During the exposure time the safeguards personnel is free for other activities and inspections, he just has to collect the sample holders after exposure. In big facilities with ample work to do this could be positive, in smaller research facilities this might be a disadvantage. In summary it has been shown that with SSTR's and TLD's excellent results can be obtained which match the results obtained by fission- and ionization chambers. The type of method which is used for safeguards inspection finally depends on the type of facility, on the fuel elements, on the equipment, and on time and space available for the inspection. (Author)

  13. A non-destructive technique for qualitative and attribute verification of spent fuel elements

    International Nuclear Information System (INIS)

    Vana, N.

    1982-08-01

    Some results of spent fuel measurements at Halden HWR (Norway) are described. Three irradiated assemblies were measured by using solid state track records (SSTR), thermo luminescent detectors (TLD) and small ion (IC) and fission chambers (FC). The burn-up range was from 5 to 40 GWd/t, cooling time varied from approximately 3.5 months to 12 years. The gross gamma and neutron burn-up profiles, obtained by using the above mentioned techniques are identical. The exposure time for SSTR's varied from 10 to 50 minutes. One should note that SSTR's were quite sensitive and allowed the authors to observe an ''enrichment'' effect: at nearly the same burn-up values (approximately 37 GWd/t), the assembly with lower U-235 enrichment (5.08%) had higher neutron emission than the assembly with 8% enrichment. A simple power function relationship between the neutron rate and burn-up were derived: neutron rate = α.(burn-up)sup(β), where α approximately equals 7.10 -14 ; β approximately equals 3.1 +- 47% for enrichment epsilon = 5.1% and α approximately equals 6.10 -12 ; β approximately equals 2.5 +- 62% for enrichment epsilon = 8%. These data are in qualitative agreement with the IAEA spent fuel measurements at Loviisa (Finland) and Kozloduy (Bulgaria). They also agree with the Los Alamos calculations of the neutron emission due mainly to the spontaneous fission of some curium isotopes: Cm-242 and Cm-244

  14. Burn of actinides in MOX fuel cells; Quemado de actinidos en celdas de combustible MOX

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.; Ramirez S, J. R.; Alonso V, G., E-mail: eduardo.martinez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    The spent fuel from nuclear reactors is stored temporarily in dry repositories in many countries of the world. However, the main problem of spent fuel, which is its high radio-toxicity in the long term, is not solved. A new strategy is required to close the nuclear fuel cycle and for the sustain ability of nuclear power generation, this strategy could be the recycling of plutonium to obtain more energy and recycle the actinides generated during the irradiation of the fuel to transmute them in less radioactive radionuclides. In this work we evaluate the quantities of actinides generated in different fuels and the quantities of actinides that are generated after their recycling in a thermal reactor. First, we make a reference calculation with a regular enriched uranium fuel, and then is changed to a MOX fuel, varying the plutonium concentrations and determining the quantities of actinides generated. Finally, different amounts of actinides are introduced into a new fuel and the amount of actinides generated at the end of the fuel burn is calculated, in order to determine the reduction of minor actinides obtained. The results show that if the concentration of plutonium in the fuel is high, then the production of minor actinides is also high. The calculations were made using the cell code CASMO-4 and the results obtained are shown in section 6 of this work. (Author)

  15. Non-destructive studies of fuel pellets by neutron resonance absorption radiography and thermal neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Tremsin, A.S., E-mail: ast@ssl.berkeley.edu [University of California, Berkeley, CA 94720 (United States); Vogel, S.C.; Mocko, M.; Bourke, M.A.M.; Yuan, V.; Nelson, R.O.; Brown, D.W. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Feller, W.B. [NOVA Scientific, Inc., 10 Picker Rd., Sturbridge, MA 01566 (United States)

    2013-09-15

    Many isotopes in nuclear materials exhibit strong peaks in neutron absorption cross sections in the epithermal energy range (1–1000 eV). These peaks (often referred to as resonances) occur at energies specific to particular isotopes, providing a means of isotope identification and concentration measurements. The high penetration of epithermal neutrons through most materials is very useful for studies where samples consist of heavy-Z elements opaque to X-rays and sometimes to thermal neutrons as well. The characterization of nuclear fuel elements in their cladding can benefit from the development of high resolution neutron resonance absorption imaging (NRAI), enabled by recently developed spatially-resolved neutron time-of-flight detectors. In this technique the neutron transmission of the sample is measured as a function of spatial location and of neutron energy. In the region of the spectra that borders the resonance energy for a particular isotope, the reduction in transmission can be used to acquire an image revealing the 2-dimensional distribution of that isotope within the sample. Provided that the energy of each transmitted neutron is measured by the neutron detector used and the irradiated sample possesses neutron absorption resonances, then isotope-specific location maps can be acquired simultaneously for several isotopes. This can be done even in the case where samples are opaque or have very similar transmission for thermal neutrons and X-rays or where only low concentrations of particular isotopes are present (<0.1 atom% in some cases). Ultimately, such radiographs of isotope location can be utilized to measure isotope concentration, and can even be combined to produce three-dimensional distributions using tomographic methods. In this paper we present the proof-of-principle of NRAI and transmission Bragg edge imaging performed at Flight Path 5 (FP5) at the LANSCE pulsed, moderated neutron source of Los Alamos National Laboratory. A set of urania mockup

  16. Safety evaluation on MOX new fuel at marine transport

    International Nuclear Information System (INIS)

    Tsumune, Daisuke; Ito, Chihiro; Saegusa, Toshiari; Maruyama, Koki

    2000-01-01

    In the Central Research Institute of Electric Power Industry, in order to confirm effects of MOX new fuel on the public are as small as possible even when its marine transport goes down, some exposed radiation dose has previously conducted on imaginary shipwreck of marine transport on used nuclear fuel, plutonium dioxide, and high level return glass solid. Under a base of such informations, some investigations on safety on marine transport of the MOX new fuel was conducted. On September, 1999, five transport vessels of the MOX new fuel was at first transported on marine. The value of five times of estimated exposed radiation dose (max. 8.1 x 10 -8 mSv/y) corresponds to an evaluation result assumed by shipwreck in marine transport this time. As a result, it was found that the exposed radiation dose estimated on this case would be sufficiently less than an effective dose equivalent limit (1 mSv/y) of public exposure according to the recommendation of ICRP in both coastal and oceanic areas. (G.K.)

  17. Safety-related investigations on power distribution in MOX fuel elements in LWR cores

    International Nuclear Information System (INIS)

    Kramer, E.; Langenbuch, S.

    1991-01-01

    For the concept of thermal recycling various fuel assembly designs have been developped during the last years. An overview is given describing the present status of MOX-fuel assembly design for PWR and BWR. The local power distribution within the MOX-fuel assembly and influences between neighbouring MOX- and Uranium fuel assemblies have been analyzed by own calculations. These investigations are limited to specific aspects of the spatial power distribution, which are related to the use of MOX-fuel assemblies within the reactor core of LWR. (orig.) [de

  18. Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, Ronald James [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-01

    The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) during cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.

  19. MOX fuel use as a back-end option: Trends, main issues and impacts on fuel cycle management

    International Nuclear Information System (INIS)

    Fukuda, K.; Choi, J.-S.; Shani, R.; Durpel, L. van den; Bertel, E.; Sartori, E.

    2000-01-01

    In the past decades while the FBIULWR fuel cycle concept was zealously being developed, MOX-fuel use in thermal reactors was taken as an alternative back-end policy option. However, the plutonium recycling with LWRs has evolved to industrial level, gaining high maturity through the incubative period while FBR deployment was envisaged. Today, MOX-fuel use in LWRs makes integral part of the fuel cycle for those countries relying on the recycling policy. Developments to improve the fuel cycle performance, including the minimisation of remaining wastes, and the reactor engineering aspects owing to MOX-fuel use, are continued. This paper jointly presented by IAEA and OECD/NEA brings an integrated overview on MOX use as a back-end policy, covering MOX fuel utilisation, fuel performance and technology, economics, licensing, MOX fuel trends in the coming decades. (author)

  20. An overview of economic and technical issues related to LWR MOX fuel usage

    International Nuclear Information System (INIS)

    Malone, J.P.; Varley, G.; Goldstein, L.

    1999-01-01

    This paper will present comparisons of the economics of MOX versus UO 2 fuels. In addition to the economics of the front end, the scope of the comparison will include the back end of the fuel cycle. Management of spent MOX fuel assemblies presents utilities with some technical issues that can complicate spent fuel pool operation. Alternative spent fuel management methods, such as dry storage of spent MOX fuel assemblies, will also be discussed. Differences in decay heat loads versus time for spent MOX and UO 2 fuel assemblies will be presented. This difference is one of the main problems confronting spent fuel managers relative to MOX. The difference in decay heat loads will serve as the basis for a performance overview of the various spent fuel technologies available today. The economics of the front end of MOX will be presented relative to UO 2 fuel. Availability of MOX manufacturing capability will also be discussed, along with a discussion of its impact on future MOX fabrication prices. The in-core performance of MOX will be compared to that of UO 2 fuel with similar performance characteristics. The information will include highlights of nuclear design and related operational considerations such as: Reactivity reduction with burnup is slower for MOX fuel than for UO 2 fuel; Spectral hardening resulting in lower control rod worths and a lower soluble boron worth; and more negative moderator, void and fuel temperature coefficients. A comparison of Westinghouse and ABB-CE core designs for use on disposition of weapons MOX in 12- and 18-month cycles will be presented. (author)

  1. The nuclear fuel: Mox and Melox

    International Nuclear Information System (INIS)

    Bekiarian, A.; Nigon, J.L.

    1991-01-01

    In this paper we indicate the policy used in France for the utilization of plutonium in the PWR and we give the actual state of MELOX facility construction and the schema of fuel assembly manufacturing [fr

  2. Microwave processing in MOX fuel cycle

    International Nuclear Information System (INIS)

    Mallik, G.K.; Malav, R.K.; Panakkal, J.P.; Kamath, H.S.

    2005-01-01

    The prominent aspect of the microwave heating technique applications in nuclear material processing is its eco-friendly status. It is envisaged that no active liquid waste will be generated from microwave processing. AFFF has fabricated the (U, Pu) 2 O mixed oxide fuels for PHWRs, BWRs and PFBR. AFFF is also working for the AHWR fuel cycle. The present paper summarises about the process experiments, instrumental development, results, and future applications of microwave heating technique. (author)

  3. A comparative study of fission gas behaviour in UO2 and MOX fuels using the meteor fuel performance code

    International Nuclear Information System (INIS)

    Struzik, C.; Garcia, Ph.; Noirot, L.

    2002-01-01

    The paper reviews some of the fission-gas-related differences observed between MOX MIMAS AUC fuels and homogeneous UO 2 fuels. Under steady-state conditions, the apparently higher fractional release in MOX fuels is interpreted with the METEOR fuel performance code as a consequence of their lower thermal conductivity and the higher linear heat rates to which MOX fuel rods are subjected. Although more fundamental diffusion properties are needed, the apparently greater swelling of MOX fuel rods at higher linear heat rates can be ascribed to enhanced diffusion properties. (authors)

  4. Novel technique for manipulating MOX fuel particles using radiation pressure of a laser light

    International Nuclear Information System (INIS)

    Omori, R.

    2000-01-01

    We have continued theoretical and experimental studies on laser manipulation of nuclear fuel particles, such as UO 2 , PuO 2 and ThO 2 , In this paper, we investigate the applicability of the collection of MOX particles floating in air using radiation pressure of a laser light; some preliminary results are shown. This technique will be useful for removal and confinement of MOX particles being transported by air current or dispersed in a cell box. First, we propose two types of principles for collecting MOX particles. Second, we show some experimental results, Third, we show numerical results of radiation pressure exerted on submicrometer-sized UO 2 particles using Generalized Lorentz-Mie theory. Because optical constants of UO 2 are similar to those of MOX fuel particles, it seems that calculation results obtained hold for MOX fuel particles. 2. Principles of collecting MOX fuel particles using radiation pressure (authors)

  5. Implications of plutonium and americium recycling on MOX fuel fabrication

    International Nuclear Information System (INIS)

    Renard, A.; Pilate, S.; Maldague, Th.; La Fuente, A.; Evrard, G.

    1995-01-01

    The impact of the multiple recycling of plutonium in power reactors on the radiation dose rates is analyzed for the most critical stage in a MOX fuel fabrication plant. The limitation of the number of Pu recycling in light water reactors would rather stem from reactor core physics features. The case of recovering americium with plutonium is also considered and the necessary additions of shielding are evaluated. A comparison between the recycling of Pu in fast reactors and in light water reactors is presented. (author)

  6. Mixed-oxide (MOX) fuel performance benchmark. Summary of the results for the PRIMO MOX rod BD8

    International Nuclear Information System (INIS)

    Ott, L.J.; Sartori, E.; Costa, A.; ); Sobolev, V.; Lee, B-H.; Alekseev, P.N.; Shestopalov, A.A.; Mikityuk, K.O.; Fomichenko, P.A.; Shatrova, L.P.; Medvedev, A.V.; Bogatyr, S.M.; Khvostov, G.A.; Kuznetsov, V.I.; Stoenescu, R.; Chatwin, C.P.

    2009-01-01

    The OECD/NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, nuclear fuel performance, and fuel cycle issues related to the disposition of weapons-grade plutonium as MOX fuel. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close cooperation with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A major part of these activities includes benchmark studies. This report describes the results of the PRIMO rod BD8 benchmark exercise, the second benchmark by the TFRPD relative to MOX fuel behaviour. The corresponding PRIMO experimental data have been released, compiled and reviewed for the International Fuel Performance Experiments (IFPE) database. The observed ranges (as noted in the text) in the predicted thermal and FGR responses are reasonable given the variety and combination of thermal conductivity and FGR models employed by the benchmark participants with their respective fuel performance codes

  7. Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J.

    2005-05-27

    The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.

  8. Results of Am isotopic ratio analysis in irradiated MOX fuels

    Energy Technology Data Exchange (ETDEWEB)

    Koyama, Shin-ichi; Osaka, Masahiko; Mitsugashira, Toshiaki; Konno, Koichi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Kajitani, Mikio

    1997-04-01

    For analysis of a small quantity of americium, it is necessary to separate from curium which has similar chemical property. As a chemical separation method for americium and curium, the oxidation of americium with pentavalent bismuth and subsequent co-precipitation of trivalent curium with BIP O{sub 4} were applied to analyze americium in irradiated MOX fuels which contained about 30wt% plutonium and 0.9wt% {sup 241}Am before irradiation and were irradiated up to 26.2GWd/t in the experimental fast reactor Joyo. The purpose of this study is to measure isotopic ratio of americium and to evaluate the change of isotopic ratio with irradiation. Following results are obtained in this study. (1) The isotopic ratio of americium ({sup 241}Am, {sup 242m}Am and {sup 243}Am) can be analyzed in the MOX fuels by isolating americium. The isotopic ratio of {sup 242m}Am and {sup 243}Am increases up to 0.62at% and 0.82at% at maximum burnup, respectively, (2) The results of isotopic analysis indicates that the contents of {sup 241}Am decreases, whereas {sup 242m}Am, {sup 243}Am increase linearly with increasing burnup. (author)

  9. Hot vacuum outgassing to ensure low hydrogen content in MOX fuel pellets for thermal reactors

    International Nuclear Information System (INIS)

    Majumdar, S.; Nair, M.R.; Kumar, Arun

    1983-01-01

    Hot vacuum outgassing treatment to ensure low hydrogen content in Mixed Oxide Fuel (MOX) pellets for thermal reactors has been described. Hypostoichiometric sintered MOX pellets retain more hydrogen than UO 2 pellets. The hydrogen content further increases with the addition of admixed lubricant and pore formers. However, low hydrogen content in the MOX pellets can be ensured by a hot vacuum outgassing treatment at a temperature between 773K to 823K for 2 hrs. (author)

  10. ORIGEN-2 libraries based on JENDL-3.2 for PWR-MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Hideki; Onoue, Masaaki; Tahara, Yoshihisa [Mitsubishi Heavy Industries Ltd., Tokyo (Japan)

    2001-08-01

    A set of ORIGEN-2 libraries for PWR MOX fuel was developed based on JENDL-3.2 in the Working Group on Evaluation of Nuclide Production, Japanese Nuclear Data Committee. The calculational model generating ORIGEN-2 libraries of PWR MOX is explained here in detail. The ORIGEN-2 calculation with the new ORIGEN-2 MOX library can predict the nuclides contents within 10% for U and Pu isotopes and 20% for both minor actinides and main FPs. (author)

  11. Transportation and packaging issues involving the disposition of surplus plutonium as MOX fuel in commercial LWRs

    International Nuclear Information System (INIS)

    Ludwig, S.B.; Welch, D.E.; Best, R.E.; Schmid, S.P.

    1997-08-01

    This report provides a view of anticipated transportation, packaging, and facility handling operations that are expected to occur at mixed-oxide (MOX) fuel fabrication and commercial reactor facilities. This information is intended for use by prospective contractors to the U.S. Department of Energy (DOE) who plan to submit proposals to DOE to manufacture and irradiate MOX fuel assemblies in domestic commercial light-water reactors. The report provides data to prospective consortia regarding packaging and pickup of MOX nuclear fuel assemblies at a MOX fuel manufacturing plant and transport and delivery of the MOX assemblies to nuclear power plants. The report also identifies areas where data are incomplete either because of the status of development or lack of sufficient information and specificity regarding the nuclear power plant(s) where deliveries will take place

  12. Radiative capture on $^{242}$Pu for MOX fuel reactors

    CERN Multimedia

    The use of MOX fuel (mixed-oxide fuel made of UO$_{2}$ and PuO$_{2}$) in nuclear reactors allows substituting a large fraction of the enriched Uranium by Plutonium reprocessed from spent fuel. Indeed around 66% of the plutonium from spent fuel is made of $^{239}$Pu and $^{241}$Pu, which are fissile in thermal reactors. A typical reactor of this type uses a fuel with 7% reprocessed Pu and 93% depleted U, thus profiting from both the spent fuel and the remaining $^{238}$U following the $^{235}$U enrichment. With the use of such new fuel compositions rich in Pu the better knowledge of the capture and fission cross sections of the Pu isotopes becomes very important. This is clearly stated in the recent OECD NEA’s “High Priority Request List” and in the WPEC-26 “Uncertainty and target accuracy assessment for innovative systems using recent covariance data evaluations” report. In particular, a new series of cross section evaluations have been recently carried out jointly by the European (JEFF) and United ...

  13. CHF considerations for highly moderated 100% MOX fuels PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Saphier, D.; Raymond, P. [CEA Saclay, DMT/SERMA/LETR, Gif-sur-Yvette (France)

    1995-09-01

    A feasibility study on using 100% MOX fuel in a PWR with increased moderating ratio, RMA, was initiated. In the proposed design all the parameters were chosen identical to the French 1450MW PWR, except the fuel pin diameter which was reduced to achieve higher moderating ratios, V{sub M}/V{sub F}, where V{sub M} and V{sub F} are the moderator and fuel volume respectively. Moderating ratios from 2 to 4 were considered. In the present study the thermal-hydraulic feasibility of using fuel assemblies with smaller diameter fuel pins was investigated. The major design constrain in this study was the critical heat flux (CHF). In order to maintain the fuel pin integrity under nominal operating and transient conditions, the minimum DNBR, (Departure from Nucleate Boiling Ratio given by CHF/q{close_quotes}{sub local}, where q{close_quotes}{sub local} is the local heat flux), has to be above a given value. The limitations of the existing CHF correlations for the present study are outlined. Two designs based on the conventional 17x17 fuel assembly and on the advanced 19x19 assembly meeting the MDNBR criteria and satisfying the control margin requirements, are proposed.

  14. Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs

    International Nuclear Information System (INIS)

    Mertyurek, Ugur; Gauld, Ian C.

    2016-01-01

    Highlights: • ORIGEN MOX library generation process is described. • SCALE burnup calculations are validated against measured MOX fuel samples from the MALIBU program. • ORIGEN MOX libraries are verified using the OECD Phase IV-B benchmark. • There is good agreement for calculated-to-measured isotopic distributions. - Abstract: ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. The nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

  15. ORIGEN2 libraries based on JENDL-3.2 for LWR-MOX fuels

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Katakura, Jun-ichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Onoue, Masaaki; Matsumoto, Hideki [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Sasahara, Akihiro [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

    2000-11-01

    A set of ORIGEN2 libraries for LWR MOX fuels was developed based on JENDL-3.2. The libraries were compiled with SWAT using the specification of MOX fuels that will be used in nuclear power reactors in Japan. The verification of the libraries were performed by the analyses of post irradiation examinations for the fuels from European PWR. By the analysis of PIE data from PWR in United States, the comparison was made between calculation and experimental results in the case of that parameters for making the libraries are different from irradiation conditions. These new libraries for LWR MOX fuels are packaged in ORLIBJ32, the libraries released in 1999. (author)

  16. Recent prospects of MOX fuel and strategy about nuclear fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1991-04-01

    It is clearly described what is the preliminary adequate strategic concern for different nuclear power countries under different nuclear power development conditions. It is also stressed on the basic situation of the design technology, manufacture technology, operation experiences and quantitative economic analysis for MOX fuel application since fast breed reactor commercialization has been delayed. The author specially proposed that in a short term China should adopt an intermediate storage strategy matched with the construction of a pilot reprocessing plant to prepare the technical basis for commercialized reprocessing plant later on and to follow the development of MOX fuel technology

  17. LWR mox fuel experience in Belgium and France with special emphasis on results obtained in BR3

    International Nuclear Information System (INIS)

    Bairiot, H.; Haas, D.; Lippens, M.; Motte, F.; Lebastard, G.; Marin, J.F.

    1986-09-01

    The course of the paper reflects two main topics: LWR MOX fuel experience in Belgium and France, summarizing the fabrication techniques, the references, the underlying MOX fuel technology and the current R and D programs for expanding the data base; behaviour of MOX fuel rods irradiated under steady state and transient operating conditions, focusing on MOX fuel technology features acquired through the irradiations performed in the BR3 PWR, supplemented by tests in the BR2 MTR. This paper focuses on the thermomechanical behaviour of LWR MOX fuel rods, which is intimately related to the fabrication technique and vice-versa. 22 refs

  18. Development of a fresh MOX fuel transport package for disposition of weapons plutonium

    International Nuclear Information System (INIS)

    Ludwig, S.B.; Pope, R.B.; Shappert, L.B.; Michelhaugh, R.D.; Chae, S.M.

    1998-01-01

    The US Department of Energy announced its Record of Decision on January 14, 1997, to embark on a dual-track approach for disposition of surplus weapons-usable plutonium using immobilization in glass or ceramics and burning plutonium as mixed-oxide (MOX) fuel in reactors. In support of the MOX fuel alternative, Oak Ridge National Laboratory initiated development of conceptual designs for a new package for transporting fresh (unirradiated) MOX fuel assemblies between the MOX fabrication facility and existing commercial light-water reactors in the US. This paper summarizes progress made in development of new MOX transport package conceptual designs. The development effort has included documentation of programmatic and technical requirements for the new package and development and analysis of conceptual designs that satisfy these requirements

  19. Main trends and content of works on fabrication of fuel rods with MOX fuel for the WWER-1000 reactor

    International Nuclear Information System (INIS)

    Tsykanov, V.A.; Golovanov, V.N.; Mayorshin, A.A.; Yurchenko, A.D.; Ilyenko, S.A.; Syuzev, V.N.

    2000-01-01

    The main trends of production of pellet MOX-fuel for the WWER reactors using the trial-experimental equipment at SSC RF RIAR are set forth. The main realized parameters of fabrication of MOX-fuel pellets are presented. The content of the reactor tests program is considered with allowance for their licensing requirements for the WWER reactors. (author)

  20. 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.

  1. Water reactor fuel characterization. Part of a coordinated programme on non-destructive techniques for reactor fuel characterization

    International Nuclear Information System (INIS)

    Levai, F.

    1983-06-01

    The report describes an optical/mechanical system of examining nuclear reactor fuel bundles by tomographic imaging using high contrast X-ray film. This low cost system does not use expensive detectors or digital computers. The apparatus assembled from ordinary and available components consists of a 2π scanner, a back projector and filters. Although the system described and tested is based on transmission tomography, the report also discusses the extension of the concept to emission tomography

  2. 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)

  3. 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.

  4. DRAGON analysis of MOX fueled VVER cell benchmarks

    International Nuclear Information System (INIS)

    Marleau, G.; Foissac, F.

    2002-01-01

    The computational unit-cell benchmarks problems for LEU and MOX fueled VVER-1000 ('water-water energetic reactor') have been analyzed using the code DRAGON with ENDF/B-V and ENDF/B-VI based WIMS-AECL cross section libraries. The results obtained were compared with those generated using the SAS2H module of the SCALE-4.3 computational code system and with the code HELIOS. Good agreements between DRAGON and HELIOS were obtained when the ENDF/B-VI based library was considered while the ENDF/B-V DRAGON results were generally closer to those obtained using SAS2H. This study was useful for the verification of the DRAGON code and confirms that HELIOS and DRAGON have a similar behavior when compatible cross sections library are used. (author)

  5. Radiation shielding calculation for the MOX fuel fabrication plant Melox

    International Nuclear Information System (INIS)

    Lee, Y.K.; Nimal, J.C.; Chiron, M.

    1994-01-01

    Radiation shielding calculation is an important engineering work in the design of the MOX fuel fabrication plant MELOX. Due to the recycle of plutonium and uranium from UO2 spent fuel reprocessing and the large capacity of production (120t HM/yr.), the shielding design requires more attention in this LWR fuel plant. In MELOX, besides several temporary storage facilities of massive fissile material, about one thousand radioactive sources with different geometries, forms, densities, quantities and Pu concentrations, are distributed through different workshops from the PuO 2 powder reception unit to the fuel assembly packing room. These sources, with or without close shield, stay temporarily in different locations, containers and glove boxes. In order to optimize the dimensions, the material and the cost of shield as well as to limit the calculation work in a reasonable engineer-hours, a calculation scheme for shielding design of MELOX is developed. This calculation scheme has been proved to be useful in consideration of the feedback from the evolutionary design and construction. The validated shielding calculations give a predictive but reliable radiation doses information. (authors). 2 figs., 10 refs

  6. IFPE/CNEA-MOX-RAMP, CNEA Power Ramp Irradiations with (PHWR) MOX Fuels

    International Nuclear Information System (INIS)

    Marino, Armando Carlos; Turnbull, J.A.

    2000-01-01

    Description: The irradiation of the first MOX nuclear fuel rods fabricated in Argentina began in 1986. These experiences were made in the HFR-Petten reactor, Holland. The six rods were fabricated in the a Facility (GAID-CNEA-Argentina). The first rod has been used for destructive pre-irradiation characterization in the KFK (Kernforschungszentrum Karlsruhe), Germany. The second one was a pathfinder for calibrating HFR systems in Petten. Two other rods included pellets doped with iodine. The first contained mostly CsI whilst the second contained elemental iodine. The concentration of iodine was intended to simulate a burn-up of 15000 MWd/ton(M). The power histories were defined from calculations performed with the BACO code. A 15 day cycle was assumed with a power history that induced PCMI during power cycling. The last high power period was maintained until stress corrosion cracking (SCC) was induced. Two further un-doped rods were used in a sub-program named BU15. Here a burn-up of 15000 MWd/ton(M) was achieved at a low power followed by a final power ramp for one of the rods. The ramp was similar to that used for the Iodine test. The HFR irradiation was conducted satisfactorily. The objective was to attempt a correspondence in behaviour between the doped rods and BU15 rods. PIE detected the presence of micro-cracks inside the cladding of the iodine doped rods. Ramping of the BU15 rod was interrupted when an increase of coolant activity was detected. After discharge, a visual inspection of the rod showed the presence of a small circular hole in the cladding. Additional PIE showed that the hole was due to a SCC failure

  7. International symposium on MOX fuel cycle technologies for medium and long-term deployment. Book of extended synopses

    International Nuclear Information System (INIS)

    1999-05-01

    The purpose of the Symposium was to provide a forum to exchange information on MOX fuel cycle technologies with focus on how past experience is being or can be used to progress further, either for facing more demanding fabrication and utilization conditions or for extending into new processing or utilization domains. Presented papers covered the following topics: Current status and prospects concerning plutonium management and MOX fuel utilization; MOX fuel fabrication technology and quality control; Fuel design, performance and testing; In-core fuel management and advanced fuel cycle options; Safety analysis, licensing and safeguards; Transportation and management of irradiated MOX fuel

  8. 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)

  9. Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

    International Nuclear Information System (INIS)

    Kosaka, Yuji; Kurematsu, Shigeru; Kitagawa, Takaaki; Suzuki, Akihiro; Terai, Takayuki

    2012-01-01

    The temperature measurements of mixed oxide (MOX) and UO 2 fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO 2 fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel. (author)

  10. A non-destructive, ultrasonic method for the determination of internal pressure and gas composition in an LWR fuel rod on-going and future programme

    International Nuclear Information System (INIS)

    Ferrandis, J.; Leveque, G.; Villard, J.

    2006-01-01

    Several possible non-destructive methods have been investigated in the past to measure the internal gas pressure e.g., measurement of 85 Kr directly, or after accumulation in the plenum by freezing with liquid nitrogen. However no satisfactory resolution to the problem has been found, so at present there is no rapid and accurate method of determining the fission gas pressure in a fuel rod without puncturing the cladding. This procedure is time-consuming and expensive and as a consequence a relatively small number of measurements are generally made compared with the number of fuel rods irradiated. In this paper it is proposed a new method for the measurement of pressure that is: Non-destructive; Non-invasive (i.e., allows re-irradiation of the measured rod); Easy to operate - directly in the reactor pool; Can be used on the critical path; Is inexpensive compared with the methods currently in use. This method is also being adapted to the on line measurement of fission gas release on fuel irradiation in research reactors. This method is based on the application of acoustic technology

  11. Hydrothermal synthesis for fabrication and reprocessing of MOX nuclear fuel

    International Nuclear Information System (INIS)

    Ohta, Suguru; Yamamura, Tomoo; Shirasaki, Kenji; Satoh, Isamu; Shikama, Tatsuo

    2011-01-01

    To improve the nuclear proliferation resistance and to minimize use of chemicals, a new reprocessing and fabrication process of 'mixed oxide' (MOX) fuel was proposed and studied by using simulated spent fuel solutions. The process is consisting of the two steps, i.e. the removal of fission product (FP) from dissolved spent fuel by using carbonate solutions (Step-1), and hydrothermal synthesis of uranium dioxides (Step-2). In Step-1, rare earth (the precipitation ratio: 90%) and alkaline earth (10-50% for Sr) as FP were removed based on their low solubility of hydroxides and carbonate salts, with uranium kept dissolved for the certain carbonate solutions of weak base (Type 2) or mixtures of relatively strong base and weak base (Type 3). In Step-2, the features of uranium dioxides UO 2+x particles, i.e. stoichiometry (x=0.05-0.2), size (0.2-3 μm) and shape (cubic, spherical, rectangular parallelpiped, etc.), were controlled, and the cesium was removed down to 40 ppm by an addition of organic additives. The decontamination factors (DF) for cesium exceeds 10 5 , whereas the total DF of all the simulated FP were as low as the order of 10 which requires future studies for removal of alkaline earth, Re and Tc etc. (author)

  12. Safety analysis of MOX fuels by fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

  13. MOX fuel cycle technologies for medium and long term deployment. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    More than thirty years of reactor experience using MOX fuel as well as the fabrication of 2000 MOX assemblies with the use of 85 t of Pu separated from spent fuel from power reactors indicates that the recycling of plutonium as MOX fuel in LWRs has become a mature industry. The number of countries engaged in plutonium recycling could be increasing in the near future, aiming for the reduction of stockpiles of separated plutonium from earlier and existing reprocessing contracts. Economic and strategic considerations are the main factors on which to base such a decision to use MOX. Transport of MOX fuel assemblies is a vital element in these recycle programmes but could have the potential to be a weak link in the chain. To avoid problems, it is essential that sufficient numbers of transport flasks of the required types, licensed for the increasing Pu contents, be made available in a timely manner to keep pace with the planned increases in fabrication rates. Despite the excellent safety records for radioactive and MOX transports over many decades, continuous attention should be drawn to establishing the transport modalities, buffer stores, secure vehicles, and transport routes, at the same time accounting for public sensitivities on radioactive transports in general and MOX transport in particular. A large number of technical presentations updated and reconfirmed the good and almost defect-free performance of MOX fuel at increasingly high burn-up levels. MOX fuel is designed to meet the same operational and safety criteria as uranium fuels under equivalent conditions. This is also confirmed by the parallel development of design codes to accommodate the special characteristics of MOX. Integral and specific parameter testing of MOX fuel in normal and off-normal operation is under way in a number of countries with particular emphasis on high burnup behaviour. Here the important contributions of the OECD/NEA Halden BWR programme should be mentioned. The reactor

  14. Study on transport safety of refresh MOX fuel. Radiation dose from package hypothetically submerged into sea

    International Nuclear Information System (INIS)

    Tsumune, Daisuke; Suzuki; Hiroshi; Saegusa, Toshiari; Maruyama, Koki; Ito, Chihiro; Watabe, Naoto

    1999-01-01

    The sea transport of fresh MOX fuel from Europe to Japan is under planning. For the structure and equipment of transport ships for fresh MOX fuels, there is a special safety standard called the INF Code of IMO (International Maritime Organization). For transport of radioactive materials, there is a safety standard stipulated in Regulations for the Safe Transport of Radioactive Material issued by IAEA (International Atomic Energy Agency). Under those code and standard, fresh MOX fuel will be transported safely on the sea. However, a dose assessment has been made by assuming that a fresh MOX fuel package might be sunk into the sea by unexpected reasons. In the both cases for a package sunk at the coastal region and for that sunk at the ocean, the evaluated result of the dose equivalent by radiation exposure to the public are far below the dose equivalent limit of the ICRP recommendation (1 mSv/year). (author)

  15. MOX fuel cycle technologies for medium and long term deployment. Proceedings

    International Nuclear Information System (INIS)

    2000-01-01

    More than thirty years of reactor experience using MOX fuel as well as the fabrication of 2000 MOX assemblies with the use of 85 t of Pu separated from spent fuel from power reactors indicates that the recycling of plutonium as MOX fuel in LWRs has become a mature industry. The number of countries engaged in plutonium recycling could be increasing in the near future, aiming for the reduction of stockpiles of separated plutonium from earlier and existing reprocessing contracts. Economic and strategic considerations are the main factors on which to base such a decision to use MOX. Transport of MOX fuel assemblies is a vital element in these recycle programmes but could have the potential to be a weak link in the chain. To avoid problems, it is essential that sufficient numbers of transport flasks of the required types, licensed for the increasing Pu contents, be made available in a timely manner to keep pace with the planned increases in fabrication rates. Despite the excellent safety records for radioactive and MOX transports over many decades, continuous attention should be drawn to establishing the transport modalities, buffer stores, secure vehicles, and transport routes, at the same time accounting for public sensitivities on radioactive transports in general and MOX transport in particular. A large number of technical presentations updated and reconfirmed the good and almost defect-free performance of MOX fuel at increasingly high burn-up levels. MOX fuel is designed to meet the same operational and safety criteria as uranium fuels under equivalent conditions. This is also confirmed by the parallel development of design codes to accommodate the special characteristics of MOX. Integral and specific parameter testing of MOX fuel in normal and off-normal operation is under way in a number of countries with particular emphasis on high burnup behaviour. Here the important contributions of the OECD/NEA Halden BWR programme should be mentioned. The reactor

  16. Preliminary analysis of a large 1600 MWe PWR core loaded with 30% MOX fuel

    International Nuclear Information System (INIS)

    Polidoro, Franco; Corsetti, Edoardo; Vimercati, Giuliano

    2011-01-01

    The paper presents a full-core 3-D analysis of the performances of a large 1600 MWe PWR core, loaded with 30% MOX fuel, in accordance with the European Utility Requirements (EUR). These requirements state that the European next generation power plants have to be designed capable to use MOX (UO 2 - PuO 2 ) fuel assemblies up to 50% of the core, together with UO 2 fuel assemblies. The use of MOX assemblies has a significant impact on key physic parameters and on safety. A lot of studies have been carried out in the past to explore the feasibility of plutonium recycling strategies by loading LWR reactors with MOX fuel. Many of these works were based on lattice codes, in order to perform detailed analyses of the neutronic characteristics of MOX assemblies. With the aim to take into account their interaction with surrounding UO 2 fuel elements, and the global effects on the core at operational conditions, an integrated approach making use of a 3-D core simulation is required. In this light, the present study adopts the state-of-art numerical models CASMO-5 and SIMULATE-3 to analyze the behavior of the core fueled with 30% MOX and to compare it with that of a large PWR reference core, fueled with UO 2 . (author)

  17. Novel technique for manipulating MOX fuel particles using radiation pressure of a laser light

    International Nuclear Information System (INIS)

    Omori, R.; Suzuki, A.

    2001-01-01

    We proposed two principles based on the laser manipulation technique for collecting MOX fuel particles floating in air. While Principle A was based on the acceleration of the MOX particles due to the radiation pressure of a visible laser light, Principle B was based on the gradient forces exerted on the particles when an infrared laser light was incident. Principle A was experimentally verified using MnO 2 particles. Numerical results also showed the possibility of collecting MOX fuel particles based on both the principles. (authors)

  18. The MOX fuel behaviour test IFA-597.4/.5. Temperature and pressure data to a burn-up of 15 MWd/kg MOX

    International Nuclear Information System (INIS)

    Takano, K.

    1999-04-01

    The behaviour of MOX fuel should be investigated in detail for more effective use in the future, especially concerning its thermal performance and fission gas release. IFA-597.4 and IFA-597.5, containing two MOX fuel rods each with a fuel centre thermocouple and a pressure transducer, have been irradiated in the Halden Reactor to study the temperature threshold of fission gas release for MOX fuel and to explore potential differences in the thermal and fission gas release behaviour between solid and hollow pellets. The two rods of MOX fuel with an initial Pu-fissile content of 6.07 percent have solid pellets and hollow pellets respectively, and with an active length of about 220 mm. The diameter of the pellets is 8.05 mm with 180μm of diametral gap to the cladding. For the purpose of the test, power ramp operation, in which estimated peak temperature of the MOX pellets increases and decreases above and below the threshold for fission gas release in UO 2 fuel, is planned every 10 MWd/kgMOX of burn-up. The first ramp operation has been successfully performed at 10 MWd/kgMOX. When the estimated peak temperature of the fuel gets close to but below the threshold of UO 2 , fission gas release was observed at around 28 kW/m of power. Densification of the MOX pellets could be estimated to about 1.2 percent for the solid pellets and about 2,3 percent for the hollow pellets from normalised internal rod pressure. After 13.5 MWd/kgMOX the average assembly power has been operated low enough to observe swelling rate of MOX fuel pellets and behaviour after significant fission gas release. The burn-up had reached 15.5 MWd/kgMOX as of the end of 1998. The target burn-up of this MOX test is 60 MWd/kgMOX (author) (ml)

  19. Continuous process of powder production for MOX fuel fabrication according to ''granat'' technology

    International Nuclear Information System (INIS)

    Morkovnikov, V.E.; Raginskiy, L.S.; Pavlinov, A.P.; Chernov, V.A.; Revyakin, V.V.; Varykhanov, V.S.; Revnov, V.N.

    2000-01-01

    During last years the problem of commercial MOX fuel fabrication for nuclear reactors in Russia was solved in a number of directions. The paper deals with the solution of the problem of creating a continuous pilot plant for the production of MOX fuel powders on the basis of the home technology 'Granat', that was tested before on a small-scale pilot-commercial batch-operated plant of the same name and confirmed good results. (authors)

  20. Programmatic and technical requirements for the FMDP fresh MOX fuel transport package

    International Nuclear Information System (INIS)

    Ludwig, S.B.; Michelhaugh, R.D.; Pope, R.B.

    1997-12-01

    This document is intended to guide the designers of the package to all pertinent regulatory and other design requirements to help ensure the safe and efficient transport of the weapons-grade (WG) fresh MOX fuel under the Fissile Materials Disposition Program. To accomplish the disposition mission using MOX fuel, the unirradiated MOX fuel must be transported from the MOX fabrication facility to one or more commercial reactors. Because the unirradiated fuel contains large quantities of plutonium and is not sufficient radioactive to create a self-protecting barrier to deter the material from theft, DOE intends to use its fleet of safe secure trailers (SSTs) to provide the necessary safeguards and security for the material in transit. In addition to these requirements, transport of radioactive materials must comply with regulations of the Department of Transportation and the Nuclear Regulatory Commission (NRC). In particular, NRC requires that the packages must meet strict performance requirements. The requirements for shipment of MOX fuel (i.e., radioactive fissile materials) specify that the package design is certified by NRC to ensure the materials contained in the packages are not released and remain subcritical after undergoing a series of hypothetical accident condition tests. Packages that pass these tests are certified by NRC as a Type B fissile (BF) package. This document specifies the programmatic and technical design requirements a package must satisfy to transport the fresh MOX fuel assemblies

  1. Performance of the MTR core with MOX fuel using the MCNP4C2 code

    International Nuclear Information System (INIS)

    Shaaban, Ismail; Albarhoum, Mohamad

    2016-01-01

    The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U 3 O 8 &PuO 2 ) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U 3 O 8 -Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U 3 O 8 -Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with 235 U and the amount of loaded 235 U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. - Highlights: • Re-cycling of the ETRR-2 reactor by MOX fuel. • Increase the number of the neutronic traps from one neutronic trap to three neutronic trap. • Calculation of the criticality safety and neutronic parameters of the ETRR-2 reactor for the U 3 O 8 -Al original fuel and the MOX fuel.

  2. MOX fuel: a contribution to disarmament. U.S. utilities' response to DOE's plutonium disposition decision

    International Nuclear Information System (INIS)

    Wallace, M.

    1997-01-01

    The author is chairman of the Nuclear Energy Institute Plutonium Disposition Working Group, which includes 11 nuclear utilities, including Ontario Hydro, and all the European fabricators of mixed oxide (MOX) fuel. A feasibility study is going on, to see if Russian or other weapons grade plutonium made into MOX fuel can be used in US, Canadian, or other power reactors. The US nuclear power industry is going through a period of change, and its primary responsibility must be the safe, reliable and economic operation of its plants. There is no current US MOX capacity, but the Europeans have have manufactured and burned over 400 tons of MOX fuel since 1963. Canada may be involved, initially through a pilot-scale experiment in NRU reactor

  3. Non-destructive controls

    International Nuclear Information System (INIS)

    Nouvet, A.

    1978-01-01

    The non-destructive controls permit, while respecting their integrity, the direct and individual examination of parts or complete objects as they are manufactured, as well as to follow the evolution of their eventual defects while in operation. The choice of control methods depends on the manufacturing process and shapes of parts, on the physical properties of their components as well as the nature, position and size of the defects which are likely to be detected. Whether it is a question of controls by means of ionizing radiation, flux of neutrons, ultrasons, acoustic source, sweating, magnetoscopy. Foucault currents, thermography, detection of leaks or non-destructive metallography, each has a limited field of application such that they are less competitive than complementary [fr

  4. MOX-fuel inherent proliferation-protection due to {sup 231}Pa admixture

    Energy Technology Data Exchange (ETDEWEB)

    Kryuchkov, E.F.; Glebov, V.B.; Apse, V.A.; Shmelev, A.N. [Moscow Engineering Physics Institute (State University), Moscow (Russian Federation)

    2003-07-01

    The proliferation protection levels of MOX-fuel containing small additions of protactinium are evaluated for equilibrium closed and open cycles of a light-water reactor (LWR).Analysis of the ways to the proliferation protection of MOX-fuel by small {sup 231}Pa addition and comparison of this way with another options for giving MOX-fuel the proliferation self-protection property enable us to make the 3 following conclusions: -1) Unique nature of protactinium as a small addition to MOX-fuel is determined by the following properties: - Protactinium is available in the nature (uranium ore) as a long-lived mono-isotope {sup 231}Pa, - under neutron irradiation, {sup 231}Pa is converted into {sup 232}U, which is a long-term source of high energy gamma-radiation and practically non-separable from main fuel mass, - essentially, {sup 231}Pa is a high-quality burnable neutron absorber. -2) From the proliferation self-protection point of view, nuclear fuel cycle closure with fuel recycle is a preferable option because, under this condition, introduction of protactinium into MOX-fuel allows to create the inherent radiation barrier which is in action during full cycle of fuel management at the level corresponding to the accepted today criterion of the Spent Fuel Standard (SFS). In particular, the considered example of multiple MOX-fuel recycle with small addition of {sup 231}Pa (0.2% HM) at each cycle demonstrates a possibility to reach the proliferation protection level of fresh MOX-fuel corresponding to once irradiated fuel with the same cooling time. In this case, the lethal dose (at 30 cm distance from fuel assembly) is received within the minute range. -3) Introduction of {sup 231}Pa into MOX-fuel composition in amount of 0.5% HM allows to prolong action of the SFS from 100 to 200 years. If {sup 231}Pa content is increased up to 5% HM, then MOX-fuel conserves the proliferation self-protection property in respect to short-term unauthorized actions for 200-year period of its

  5. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    O'Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO 2 and UO 2 ), typically containing 95% or more UO 2 . DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO 2 powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO 2 powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required

  6. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

  7. The need for integral critical experiments with low-moderated MOX fuels

    International Nuclear Information System (INIS)

    2004-01-01

    The use of MOX fuel in commercial reactors is a means of burning plutonium originating from either surplus weapons or reprocessed irradiated uranium fuel. This requires the fabrication of MOX assemblies on an industrial scale. The OECD/NEA Expert Group on Experimental Needs for Criticality Safety has highlighted MOX fuel manufacturing, as an area in which there is a specific need for additional experimental data for validation purposes. Indeed, integral experiments with low-moderated MOX fuel are either scarce or not sufficiently accurate to provide an appropriate degree of validation of nuclear data and computer codes. New and accurate experimental data would enable a better optimisation of the fabrication process by decreasing the uncertainties in the determination of multiplication factors of configurations such as the homogenization of MOX powders. In this context, the OECD/NEA Nuclear Science Committee organised a workshop to address the following topics: expression and justification of the need for critical or near-critical experiments employing low-moderated MOX fuels; proposals for experimental programmes to address these needs; prospects for an international co-operative programme. The workshop was held at OECD headquarters in Paris on 14-15 April 2004. (author)

  8. Pu-rich MOX agglomerate-by-agglomerate model for fuel pellet burnup analysis

    International Nuclear Information System (INIS)

    Chang, G.S.

    2004-01-01

    In support of potential licensing of the mixed oxide (MOX) fuel made from weapons-grade (WG) plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The WG-MOX comprises five percent PuO 2 and 95% depleted UO 2 . Based on the Post Irradiation Examination (PIE) observation, the volume fraction (VF) of MOX agglomerates in the fuel pellet is about 16.67%, and PuO 2 concentration of 30.0 = (5 / 16.67 x 100) wt% in the agglomerate. A pressurized water reactor (PWR) unit WG-MOX lattice with Agglomerate-by-Agglomerate Fuel (AbAF) modeling has been developed. The effect of the irregular agglomerate distribution can be addressed through the use of the Monte Carlo AbAF model. The AbAF-calculated cumulative ratio of Agglomerate burnup to U-MAtrix burnup (AG/MA) is 9.17 at the beginning of life, and decreases to 2.88 at 50 GWd/t. The MCNP-AbAF-calculated results can be used to adjust the parameters in the MOX fuel fission gas release modeling. (author)

  9. MOX Cross-Section Libraries for ORIGEN-ARP

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2003-01-01

    The use of mixed-oxide (MOX) fuel in commercial nuclear power reactors operated in Europe has expanded rapidly over the past decade. The predicted characteristics of MOX fuel such as the nuclide inventories, thermal power from decay heat, and radiation sources are required for design and safety evaluations, and can provide valuable information for non-destructive safeguards verification activities. This report describes the development of computational methods and cross-section libraries suitable for the analysis of irradiated MOX fuel with the widely-used and recognized ORIGEN-ARP isotope generation and depletion code of the SCALE (Standardized Computer Analyses for Licensing Evaluation) code system. The MOX libraries are designed to be used with the Automatic Rapid Processing (ARP) module of SCALE that interpolates appropriate values of the cross sections from a database of parameterized cross-section libraries to create a problem-dependent library for the burnup analysis. The methods in ORIGEN-ARP, originally designed for uranium-based fuels only, have been significantly upgraded to handle the larger number of interpolation parameters associated with MOX fuels. The new methods have been incorporated in a new version of the ARP code that can generate libraries for low-enriched uranium (LEU) and MOX fuel types. The MOX data libraries and interpolation algorithms in ORIGEN-ARP have been verified using a database of declared isotopic concentrations for 1042 European MOX fuel assemblies. The methods and data are validated using a numerical MOX fuel benchmark established by the Organization for Economic Cooperation and Development (OECD) Working Group on burnup credit and nuclide assay measurements for irradiated MOX fuel performed as part of the Belgonucleaire ARIANE International Program

  10. Development of recycling processes for clean rejected MOX fuel pellets

    International Nuclear Information System (INIS)

    Khot, P.M.; Singh, G.; Shelke, B.K.; Surendra, B.; Yadav, M.K.; Mishra, A.K.; Afzal, Mohd.; Panakkal, J.P.

    2014-01-01

    Highlights: • Dry and wet (MWDD) methods were developed for 100% recycling of CRO (0.4–44% PuO 2 ). • Dry method showed higher productivity and comparable powder/product characteristics. • MWDD batches demonstrated improved powder/product characteristics to that of virgin. • Second/multiple recycling is possible with MWDD with better powder/product characteristics. • MWDD batches prepared by little milling showed better macroscopic homogeneity to that of virgin. - Abstract: The dry and wet recycling processes have been developed for 100% recycling of Clean Reject Oxide (CRO) generated during the fabrication of MOX fuel, as CRO contains significant amount of plutonium. Plutonium being strategic material need to be circumvented from its proliferation issues related to its storage for long period. It was difficult to recycle CRO containing higher Pu content even with multiple oxidation and reduction steps. The mechanical recycling comprising of jaw crushing and sieving has been coupled with thermal pulverization for recycling CRO with higher Pu content in dry recycling technique. In wet recycling, MicroWave Direct Denitration (MWDD) technique has been developed for 100% recycling of CRO. The powder prepared by dry and wet (MWDD) recycling techniques was characterized by XRD and BET techniques and their effects on the pellets were evaluated. (U,21%Pu)O 2 pellets fabricated from virgin powder and MWDD were characterized using optical microscopy and α-autoradiography and the results obtained were compared

  11. Certification testing of the MOX Fresh Fuel Package (MFFP)

    International Nuclear Information System (INIS)

    Nichols, J.C. III; Yapuncich, F.L.

    2004-01-01

    Packaging Technology, Inc. (PacTec) is designing the MFFP as part of the Duke, COGEMA, Stone and Webster (DCS) consortium. DCS is tasked with providing the Department of Energy (DOE) with domestic MOX fuel fabrication and reactor irradiation services for the purpose of disposing of surplus weapons usable plutonium. This paper will discuss the development of the MFFP certification test program. The MFFP was subjected to a total of eleven free and puncture drops of the course of the certification testing. Because of the plutonium content, the design must be a Type BF, which among other things requires a containment boundary with a tested leakage rate of 1 x 10 -7 cm 3 /s air at 1 atm absolute and 25 C, or less. Both economics (desire for maximized payload) and operational (conveyance mode restricts size and weight) constraints lead to a highly optimized design. The optimized package design led to a significant test program which needed to address the containment boundary stability, puncture resistance of the package and lid end impact limiter, structural performance of the light weight lid structure, and stability of the internal structures. The test program efficiently balanced the test objectives while minimizing the number of costly hardware items used during this destructive testing. This balance achieved by strategic replacement of mock and prototypic payloads, impact limiters, and by careful test order considerations. The paper will conclude with a selected summary of the testing and an assessment of the test programs thoroughness

  12. Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

    International Nuclear Information System (INIS)

    Ozdemir, Levent; Acar, Banu Bulut; Zabunoglu, Okan H.

    2011-01-01

    When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of 239 Pu and 241 Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

  13. The development of B.N.F.L.'S MOX fuel supply business

    International Nuclear Information System (INIS)

    Edwards, J.; Brown, C.; Marshall, S.J.; Connell, M.; Thompson, H.

    1998-01-01

    In 1990 BNFL developed a strategy to become one of the world leading MOX fuel suppliers. This strategy involved the design, construction and operation of a small scale demonstration plant known as the MOX Demonstration Facility (MDF) and a large scale facility known as the Sellafield MOX Plant (SMP). To support the development of these facilities, BNFL developed a new MOX fuel fabrication process known as the Short Binderless Route (SBR). Since the 1990 decision was made, the company has successfully built, commissioned and operated the MDF, and has designed, built and is in the process of commissioning the 120 t(HM)/year SMP. The scale of the business has thus developed significantly and the direction and prospects for the future of the business are clear and well understood, with the focus being on the use of BNFL technology to produce quality MOX fuel to meet customers' requirements. This paper reviews the development of BNFL's MOX business and describes the technology being used in the state of the art SMP. The paper also explains the approach taken to commission the plant and how key safety features have been incorporated into the design. Up to date information on the performance of Short Binderless Route fuel is provided, and the future development of the business is discussed. (author)

  14. The MOX Fuel Behaviour Test IFA-597.4: Temperature And Pressure Data To A Burn-Up Of 5.4 MWd/kg MOX

    International Nuclear Information System (INIS)

    McGrath, M. A.; Teshima, H.

    1998-02-01

    Characterising the behaviour of MOX fuel is becoming increasingly important as many commercial reactors are or will be operating with this type of fuel. With this as a driving force, a new joint programme experiment, IFA-597.4, has been loaded into the reactor at Halden for the purpose of establishing the fission gas release behaviour of MOX fuel. Both annular and solid pellet fuel is being utilised and the irradiation is being conducted such that the fuel is initially operated below the onset of fission gas release. The fuel will later be subjected to small power up ratings which will be held for short periods of time. These are designed to bring the fuel to just above the temperature threshold for fission gas release thus allowing the FGR behaviour of both solid and annular MOX fuel to be established. The rig contains two fuel rods of active length 220 mm and diameter 8.05 mm. Both fuel rods contain MOX fuel with an initial Pu-fissile content of 6.07% and both are instrumented with a fuel centre thermocouple and a pressure transducer. The test is being performed under HBWR conditions and at the time of the reactor shutdown at the end of 1997 a mean burn-up of 5.4 MWd/kg MOX had been achieved with the rods at an average rating of 30 kW/m. The rod pressure data show that no fission gas had been released up to the shutdown. The fuel centre temperatures of both rods exhibit an initial increase concurrent with a fall in the monitored rod internal pressures as a result of fuel densification. It was estimated that about 1-1.4% fuel densification by volume had occurred in the two rods by a burn-up of about 3 MWd/kg MOX. (author)

  15. MOX fuel irradiation behaviour: Results from X-ray microbeam analysis

    International Nuclear Information System (INIS)

    Walker, C.T.; Goll, W.; Matsumura, T.

    1997-01-01

    The behaviour of plutonium, xenon and caesium were investigated in two sections of irradiated MOX fuel produced by the OCOM process. In one fuel (OCOM30), the MOX agglomerates contained 18 wt% fissile plutonium, and had a low volume fraction of 0.17; in the other (OCOM15) the agglomerates contained 9 wt% fissile plutonium, and had a high volume fraction of 0.34. Both fuels had been irradiated under normal power reactor conditions to a burn-up of approximately 44 GWd/t. The main aim of the work was to establish whether the above differences in composition affected the percentage fission gas released by the fuels. Since U/Pu interdiffusion did not occurred during the irradiation, both fuels remained inhomogeneous on the microscopic scale. However, the concentration of plutonium in the MOX agglomerates decreases by about 50% as a result of fission, whereas the plutonium content of the UO 2 matrix increased by about a factor of four to approximately 2 wt% due to neutron capture by 238 U. The agglomerates in the OCOM15 fuel generally exhibited a finer structure due to the lower burn-up. More than 80% of the fission gas had been released from the oxide lattice of the MOX agglomerates in both fuels. However, a very high fraction of this gas precipitated and remained in the pore structure of the agglomerates. Consequently, puncturing revealed that for both fuels the percentage of gas released to the rod free volume increased from less than 0.5% at 10 GWd/t to a maximum of 3.5% at 45 GWd/t. The conclusion is that the percentage of gas released by MOX fuel is largely unaffected of the level of inhomogeneity of the fuel. In both fuels caesium showed near complete retention in both the MOX agglomerates and the UO 2 matrix. (author). 8 refs, 11 figs, 3 tabs

  16. Simulation of facility operations and materials accounting for a combined reprocessing/MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Coulter, C.A.; Whiteson, R.; Zardecki, A.

    1991-01-01

    We are developing a computer model of facility operations and nuclear materials accounting for a facility that reprocesses spent fuel and fabricates mixed oxide (MOX) fuel rods and assemblies from the recovered uranium and plutonium. The model will be used to determine the effectiveness of various materials measurement strategies for the facility and, ultimately, of other facility safeguards functions as well. This portion of the facility consists of a spent fuel storage pond, fuel shear, dissolver, clarifier, three solvent-extraction stages with uranium-plutonium separation after the first stage, and product concentrators. In this facility area mixed oxide is formed into pellets, the pellets are loaded into fuel rods, and the fuel rods are fabricated into fuel assemblies. These two facility sections are connected by a MOX conversion line in which the uranium and plutonium solutions from reprocessing are converted to mixed oxide. The model of the intermediate MOX conversion line used in the model is based on a design provided by Mike Ehinger of Oak Ridge National Laboratory (private communication). An initial version of the simulation model has been developed for the entire MOX conversion and fuel fabrication sections of the reprocessing/MOX fuel fabrication facility, and this model has been used to obtain inventory difference variance estimates for those sections of the facility. A significant fraction of the data files for the fuel reprocessing section have been developed, but these data files are not yet complete enough to permit simulation of reprocessing operations in the facility. Accordingly, the discussion in the following sections is restricted to the MOX conversion and fuel fabrication lines. 3 tabs

  17. Evaluation of the characteristics of uranium and plutonium Mixed Oxide (MOX) fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    MOX fuel irradiation test up to high burnup has been performed for five years. Irradiation test of MOX fuel having high plutonium content has also been performed from JFY 2007 and it still continues. A lot of irradiation data have been obtained through these tests. The activities done in JFY 2012 are mainly focused on Post Irradiation Examination (PIE) data analysis concerning thermal property change and fission gas release. In the former work thermal conductivity degradation due to burnup is examined and in the latter work the dependence of fission gas release mechanism on fuel pellet microstructure is examined. This report mainly covers the result of analysis. It is found that thermal conductivity degradation of MOX fuel due to burnup is less than that of UO{sub 2} fuel and that fission gas release mechanism of high enriched fissile zone (so called Pu spot) is much different from that of low enriched fissile zone (so called Matrix). (author)

  18. Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

    International Nuclear Information System (INIS)

    Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

    2011-01-01

    The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.

  19. Implementation of the non-destructive ultrasound testing by immersion through the transmission technique, applied to the quality control of nuclear fuel plates

    International Nuclear Information System (INIS)

    Medina Jofre, David Christian

    2014-01-01

    Within the framework of global development, which seeks to reduce the enrichment of U 235 in nuclear fuels for research reactors, the Fuel Elements Plant (PEC) of the Chilean Nuclear Energy Commission (CCHEN) has worked with the Idaho National Laboratory (INL-USA), for the fabrication of high density fuel plates based on the dispersion of Uranium-Molybdenum alloy powders (UMo), which are subjected to inspections and tests to qualify as a compliant product for use in nuclear research reactors. It is in this matter where the Non Destructive Test (NDT) of immersion ultrasound used in both facilities differs in its acceptance criteria, when is used different testing techniques; On the one hand, the PEC uses the pulse-echo technique, while the INL uses the transmission technique. Therefore, the present work is focused on the implementation of the ultrasound by immersion using the transmission technique. During the development of the work, the physical and virtual configuration of the ultrasound equipment was possible and elaborate an operation procedure, which allows to inspect through this technique, a series of fuel plates based on UMo and U 3 Si 2 powders, with different characteristics. The results allow to characterize the signals obtained in fuel plates according to the nuclear fuel material used. There is an inverse relationship between the uranium load per unit volume (uranium density, gU/cm 3 ) used in the fuel plate and the transmittance of the ultrasonic beam through the areas where there is nuclear fuel material (meat); the effect produced by a dispersed combustible material is observed and it is possible to identify discontinuities that may be present in the fuel plate. Finally, an inspection technical instruction for U 3 Si 2 fuel plates is elaborated, where acceptance and rejection criteria are defined

  20. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    International Nuclear Information System (INIS)

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O'Connor, D.G.; Carrell, R.D.; Jaeger, C.D.; Thompson, M.L.; Strasser, A.A.

    1998-03-01

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET

  1. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O`Connor, D.G. [Oak Ridge National Lab., TN (United States); Carrell, R.D. [Technical Resources International, Inc., Richland, WA (United States); Jaeger, C.D. [Sandia National Labs., Albuquerque, NM (United States); Thompson, M.L.; Strasser, A.A. [Delta-21 Resources, Inc., Oak Ridge, TN (United States)

    1998-03-01

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET.

  2. Nuclear design for high temperature gas cooled reactor (GTHTR300C) using MOX fuel

    International Nuclear Information System (INIS)

    Mouri, Tomoaki; Kunitomi, Kazuhiko

    2008-01-01

    A design study of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C) that can produce both electricity and hydrogen has been carried out in Japan Atomic Energy Agency. The GTHTR300C is the system with thermal power of 600MW and reactor outlet temperature of 950degC, which is expected to supply the hydrogen to fuel cell vehicles after 2020s. In future, the full deployment of fast reactor cycle without natural uranium will demand the use of Mixed-Oxide (MOX) fuels in the GTHTR300C. Therefore, a nuclear design was performed to confirm the feasibility of the reactor core using MOX fuels. The designed reactor core has high performance and meets safety requirements. In this paper, the outline of the GTHTR300C and the nuclear design of the reactor core using MOX fuels are described. (author)

  3. Influence of plutonium contents in MOX fuel on destructive forces at fuel failure in the NSRR experiment

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Jinichi; Sugiyama, Tomoyuki; Nakamura, Takehiko; Kanazawa, Toru; Sasajima, Hideo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    In order to confirm safety margins of the Mixed Oxide (MOX) fuel use in LWRs, pulse irradiation tests are planned in the Nuclear Safety Research Reactor (NSRR) with the MOX fuel with plutonium content up to 12.8%. Impacts of the higher plutonium contents on safety of the reactivity-initiated-accident (RIA) tests are examined in terms of generation of destructive forces to threat the integrity of test capsules. Pressure pulses would be generated at fuel rod failure by releases of high pressure gases. The strength of the pressure pulses, therefore, depends on rod internal - external pressure difference, which is independent to plutonium content of the fuel. The other destructive forces, water hammer, would be generated by thermal interaction between fuel fragments and coolant water. Heat flux from the fragments to the water was calculated taking account of changes in thermal properties of MOX fuels at higher plutonium contents. The results showed that the heat transfer from the MOX fuel would be slightly smaller than that from UO{sub 2} fuel fragments at similar size in a short period to cause the water hammer. Therefore, the destructive forces were not expected to increase in the new tests with higher plutonium content MOX fuels. (author)

  4. Analysis of Core Physics Experiments on Irradiated BWR MOX Fuel in REBUS Program

    International Nuclear Information System (INIS)

    Yamamoto, Toru; Ando, Yoshihira; Hayashi, Yamato

    2008-01-01

    As part of analyses of experimental data of a critical core containing a irradiated BWR MOX test bundle in the REBUS program, depletion calculations was performed for the BWR MOX fuel assemblies from that the MOX test rods were selected by using a general purpose neutronics code system SRAC. The core analyses were carried out using SRAC and a continuous energy Monte Carlo code MVP. The calculated k eff s were compared with those of the core containing a fresh MOX fuel bundle in the program. The SRAC-diffusion calculation underestimates k eff s of the both cores by 1.0 to 1.3 %dk and the k eff s of MVP are 1.001. The difference in k eff between the irradiated BWR MOX test bundle core and the fresh MOX one is 0.4 %dk in the SRAC-diffusion calculation and 0.0 %dk in the MVP calculation. The calculated fission rate distributions are in good agreement with the measurement in the SRAC-diffusion and MVP calculations. The calculated neutron flux distributions are also in good agreement with the measurement. The calculated burnup reactivity in the both calculations well reproduce the measurements. (authors)

  5. Fuel clad chemical interactions in fast reactor MOX fuels

    Energy Technology Data Exchange (ETDEWEB)

    Viswanathan, R., E-mail: rvis@igcar.gov.in

    2014-01-15

    Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements (oxygen, cesium, tellurium, iodine) in the clad-attack are discussed and many Fuel–Clad-Chemical-Interaction (FCCI) models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL (Hanford Engineering Development Laboratory) relation is recommended: d/μm = ({0.507 ⋅ [B/(at.% fission)] ⋅ (T/K-705) ⋅ [(O/M)_i-1.935]} + 20.5) for (O/M){sub i} ⩽ 1.98. A new model is proposed for (O/M){sub i} ⩾ 1.98: d/μm = [B/(at.% fission)] ⋅ (T/K-800){sup 0.5} ⋅ [(O/M){sub i}-1.94] ⋅ [P/(W cm{sup −1})]{sup 0.5}. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, (O/M){sub i} is the initial oxygen-to-(uranium + plutonium) ratio, and P is the linear power rating. For fuels with [n(Pu)/n(M = U + Pu)] > 0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.

  6. Image analysis: a tool characterising and modelling the microstructure of the MOX fuel

    International Nuclear Information System (INIS)

    Charollais, F.

    1997-01-01

    The MOX nuclear fuel, made up of about 3 to 10 % of plutonium oxide mixed with uranium oxide, is elaborated by an original manufacturing method (MIMAS process). The MOX pellets feature a singular and complex microstructure, including enriched plutonium zones dispersed in a low plutonium content matrix. Their properties as well as their performances levels are strongly linked with this microstructure. Tools, found in the literature, allowing to quantify with relevant parameters the microstructural images from different analytical equipment (optical microscopy, electron probe micro-analyser and autoradiography) have been adapted and used in order to characterize these nuclear fuels. Taking into account the heterogeneity of the MOX microstructure, we turn our's attention, at the beginning of this study, to the analysis conditions: choice of the magnification, sampling and statistical analysis of the measurements. An improvement of the ceramographic preparation of the samples, required for an automatic image analysis (of the granular structure), has been realised by thermal etching under oxidizing gas. This method enables the strong content plutonium zones to be revealed distinctly. The first part of the study concerns the characterization of the three-dimensional structure of uranium oxide and MOX fuels by average variables using the principles of mathematical morphology and stereology. The second part introduces probabilistic models, in particular the Boolean scheme, in order to improve and complete the three-dimensional characterization of the MOX fuel and more specifically the enriched plutonium islands dispersion in the pellet. [fr

  7. Non-destructive test for irradiated fuels using X-ray CT system in hot-laboratory

    International Nuclear Information System (INIS)

    Kim, Heemoon; Kim, Gil-Soo; Yoo, Boung-Ok; Tahk, Young-Wook; Cho, Moon-Sung; Ahn, Sang-Bok

    2015-01-01

    To inspect inside of irradiated fuel rod for PIE in hotcell, neutron beam and X-ray have been used. Many hot laboratories in the world have shown the results for NDT by 2-D film data. Currently, computed image processing technology instead of film has been developed and CT was applied to the X-ray and neutron beam system. In this trend, our facility needed to set up X-ray system for irradiated fuel inspection and installed in hotcell with consideration of radiation damage. In this study, X-ray system was tested to be operated with radioactive samples and was performed to inspect fuel rods and observe internal damage and dimensional change. 450kV X-ray CT system was installed in hotcell with modification and tested to check image resolution and radiation damage. The image data were analyzed by 3-D computer software. 8 fuel plates and VHTR rods were inspected and measured internal shape and dimension

  8. Dose assessment for public at the hypothetical submergence of a fresh MOX fuel package

    International Nuclear Information System (INIS)

    Tsumune, Daisuke; Saegusa, Toshiari; Suzuki, Hiroshi; Maruyama, Koki

    2000-01-01

    For the structure and equipment of transport ships for fresh MOX fuels, there is a special safety standard called the INF Code of IMO (International Maritime Organization). For transport of radioactive materials, there is a safety standard stipulated in Regulations for the Safe Transport of Radioactive Material issued by IAEA (International Atomic Energy Agency). Under those code and standard, fresh MOX fuel is transported safety on the sea. To gain the public acceptance for this transport, a dose assessment has been made by assuming that a fresh MOX fuel package might be sunk into the sea by unknown reasons. In the both cases for a package sunk at the coastal region and for that sunk at the ocean, the evaluated result of the dose equivalent by radiation exposure to the public are far below the dose equivalent limit of the ICRP recommendation (1 mSv/year). (author)

  9. Development of moderated neutron calibration fields simulating workplaces of MOX fuel facilities

    International Nuclear Information System (INIS)

    Tsujimura, Norio; Yoshida, Tadayoshi; Takada, Chie

    2005-01-01

    It is important for the MOX fuel facilities to control neutrons produced by the spontaneous fission of plutonium isotopes and those from (α,n) reactions between 18 O and α particles emitted by 238 Pu. Neutron dose meters should be calibrated for measuring these neutrons. We have developed moderated-neutron calibration fields employing a 252 Cf neutron source and moderators mainly for the characteristics evaluation and the calibration of neutron detectors used in MOX fuel facilities. Neutron energy spectrum can be adjusted by changing the position of the 252 Cf neutron source and combining different moderators to simulate the neutron field of the MOX fuel facility. This performance is realized owing to using an existing neutron irradiation room. (K. Yoshida)

  10. Nonuniform transformation field analysis of multiphase elasto viscoplastic materials: application to MOX fuels

    International Nuclear Information System (INIS)

    Roussette, S.

    2005-05-01

    The description of the overall behavior of nonlinear materials with nonlinear dissipative phases requires an infinity of internal variables. An approximate model involving only a finite number of internal variables, Nonuniform Transformation Field Analysis, is obtained by considering a decomposition of these variables on a finite set of nonuniform transformation fields, called plastic modes. The method is initially developed for incompressible elasto viscoplastic materials. Karhunen-Loeve expansion is proposed to optimize the plastic modes. Then the method is extended to porous elasto viscoplastic materials. Finally the transformation field analysis, developed by Dvorak, is applied to nuclear fuels MOX. This method enables to make sensitivity studies to determine the role of some microstructural parameters on the fuel behaviour. Moreover the adequacy of the nonuniform method for fuels MOX is shown, the final objective being to be able to apply the model to the MOX in 3D. (author)

  11. A review on the development of the MOX fuel fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, See Hyung; Lee, Yung Woo; Sohn, Dong Sung; Yang, Myung Seung; Bae, Kee Kwang; Nah, Sang Hoh; Kim, Han Soo; Lee, Jung Won; Kim, Bong Koo; Song, Keun Woo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Development of the Mixed Oxide(MOX) fuel fabrication technology was reviewed in this study. Firstly, the feasibility of Pu utilization for nuclear fuel was analyzed by comparison of nuclear characteristics between U and Pu. Secondly, the feature and problem of processes developed so far was revealed and analyzed by reviewing each process in terms of technical difficulties and in connection with the pellet characteristics. Also, fabrication facilities currently existing were analyzed to understand particularities and circumstances in view of Pu handling, and finally, in-reactor behaviors of MOX fuel was compared with those of U fuel to understand how the Pu has an effect on fuel was compared with those of U fuel to understand how the Pu has an effect on fuel pellet structure and fuel rod. 73 figs., 15 tabs., 58 refs. (Author).

  12. Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

    Energy Technology Data Exchange (ETDEWEB)

    Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

    2002-02-26

    In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment

  13. A risk-informed evaluation of MOX fuel loading in PWRS

    International Nuclear Information System (INIS)

    Lyman, E.S.

    2001-01-01

    The full text follows: The U.S. Department of Energy (DOE) has signed a contract with Duke Cogema Stone and Webster (DCS) for fabrication of mixed-oxide (MOX) fuel and irradiation of the MOX fuel at the Catawba and McGuire pressurized-water reactors (PWRs), operated by Duke Power. The first load of MOX fuel is scheduled for 2007. In order to use MOX in these plants, Duke Power will have to apply to the Nuclear Regulatory Commission (NRC) for amendments to their operating licenses. Until recently, there have been no numerical guidelines for determining the acceptability of license amendment requests. However, such guidelines are now at hand with the adoption in 1998 of NRC Regulatory Guide 1.174, which defines a maximum value for the permissible increase in risk to the public resulting from a proposed change to a nuclear plant's licensing basis (LB). The substitution of MOX fuel for low-enriched uranium (LEU) fuel in LWRs will have an impact on risk to the public that will require regulatory evaluation. One of the major differences is that use of MOX will increase the inventories of plutonium and minor actinides in the reactor core, thereby increasing the source term for certain severe accidents, such as a core melt with early containment failure or a spent fuel pool drain-down. The goal of this paper is to quantitatively evaluate the increase in risk associated with the greater actinide source term in MOX-fueled reactors, and to compare this increase with RG 1.174 guidelines. Standard computer programs (SCALE and MACCS2) are used to estimate the increase in severe accident risk to the public associated with the DCS plan to use 40% cores of weapons-grade MOX fuel. These values are then compared to the RG 1.174 acceptance criteria, using publicly available risk information. Since RG 1.174 guidelines are based on the assumption that severe accident source terms are not affected by LB changes, the RG 1.174 formalism must be modified for this case. A similar

  14. Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

    International Nuclear Information System (INIS)

    Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

    2002-01-01

    In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is

  15. Non destructive testing of PWR type fuel pencils by Foucault currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    Marchand, L.

    1984-09-01

    The device presented in this paper, has been developed to satisfy, first, the requirements of nondestructive examinations of irradiated fuel pencils. The technics used allow the acquisition of reliable and detailed information on the state of cans. The computerized data exploitation allows to integrate quickly any improvement factor while allowing to carry out the researches on the automatic diagnostic of Eddy current signals. Now, the equipments of the bench allow visual, metrological and Foucault current simultaneous examinations [fr

  16. Effect of Pu-rich agglomerate in MOX fuel on a lattice calculation

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Yamamoto, Toru; Namekawa, Masakazu

    2007-01-01

    The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) fuel on a lattice calculation has been demonstrated. The Pu-rich agglomerate parameters are defined based on the measurement data of MIMAS-MOX and the focus is on the highly enriched MOX fuel in accordance with increased burnup resulting in a higher volume fraction of the Pu-rich agglomerates. The lattice calculations with a heterogeneous fuel model and a homogeneous fuel model are performed simulating the PWR 17x17 fuel assembly. The heterogeneous model individually treats the Pu-rich agglomerate and U-Pu matrix, whereas the homogeneous model homogenizes the compositions within the fuel pellet. A continuous-energy Monte Carlo burnup code, MVP-BURN, is used for burnup calculations up to 70 GWd/t. A statistical geometry model is applied in modeling a large number of Pu-rich agglomerates assuming that they are distributed randomly within the MOX fuel pellet. The calculated nuclear characteristics include k-inf, Pu isotopic compositions, power density and burnup of the Pu-rich agglomerates, as well as the pellet-averaged Pu compositions as a function of burnup. It is shown that the effect of Pu-rich agglomerates on the lattice calculation is negligibly small. (author)

  17. MX 8: the next generation high capacity system for the transport of fresh MOX fuel

    International Nuclear Information System (INIS)

    Potelle, F.; Issard, H.

    1998-01-01

    The choice of reprocessing policy was made a long time ago in France, leading to the development of an advanced Pu recycling industry. In 1987, Saint Laurent was the first French reactor to be loaded with fresh MOX fuel. Transnucleaire, then in charge of transport packaging development, created the FS 69 concept, derived from the classical RCC concept for the transport of UO 2 fresh fuel. On the other hand, Cogema, as the main actor in the field of fuel cycle and thus in transport matters, developed the associated security truck and security caisson in order to provide the transport system with the acceptable Physical Protection devices required by French Authorities. As a whole, the security truck and the FS 69 have now been used for more than ten years with a remarkable level of efficiency and safety. Indeed, more than 600 fresh MOX fuel elements have been delivered, without any incident, both regarding safety or fuel integrity requirements. But, as a matter of fact, the replacement of FS 69 transport system is now scheduled for several reasons. First of all, the burnups achieved with UO 2 fuel progressed together with its enrichment within the last ten years, and the MOX 'equivalence' also implies that its Pu content be increased to enhance its reactor performances: from 5.25 % of Pu content today, the MOX fuel will reach 7% tomorrow, and almost 10% the day after tomorrow. Lastly, the reprocessing/recycling policy has been confirmed and amplified, leading to an increasing number of 'moxified' reactors. As a consequence, the French utility (EDF), the fuel designer (Fragema, the joint venture between Framatome and Cogema), the fuel manufacturer (Cogema), and the transporter (Transnucleaire) joined in a specific working group devoted to the development of the MX 8, the next generation high capacity system for the land transport of MOX fuel. (authors)

  18. Revised conceptual designs for the FMDP MOX fresh fuel transport package

    International Nuclear Information System (INIS)

    Ludwig, S.B.; Michelhaugh, R.D.; Shappert, L.B.; Chae, S.M.; Tang, J.S.

    1998-03-01

    The revised conceptual designs described in this document provide a foundation for the development and certification of final transport package designs that will be needed to support the disposition of surplus weapons-grade plutonium as mixed-oxide (MOX) fuel in commercial light-water reactors in the US. This document is intended to describe the revised package design concepts and summarize the results of preliminary analyses and assessments of two new concepts for fresh MOX fuel transport packages that have been developed by Oak Ridge National Laboratory during the past year in support of the Department of Energy/Office of Fissile Materials Disposition

  19. 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

  20. Discrimination of irradiated MOX fuel from UOX fuel by multivariate statistical analysis of simulated activities of gamma-emitting isotopes

    Science.gov (United States)

    Åberg Lindell, M.; Andersson, P.; Grape, S.; Hellesen, C.; Håkansson, A.; Thulin, M.

    2018-03-01

    This paper investigates how concentrations of certain fission products and their related gamma-ray emissions can be used to discriminate between uranium oxide (UOX) and mixed oxide (MOX) type fuel. Discrimination of irradiated MOX fuel from irradiated UOX fuel is important in nuclear facilities and for transport of nuclear fuel, for purposes of both criticality safety and nuclear safeguards. Although facility operators keep records on the identity and properties of each fuel, tools for nuclear safeguards inspectors that enable independent verification of the fuel are critical in the recovery of continuity of knowledge, should it be lost. A discrimination methodology for classification of UOX and MOX fuel, based on passive gamma-ray spectroscopy data and multivariate analysis methods, is presented. Nuclear fuels and their gamma-ray emissions were simulated in the Monte Carlo code Serpent, and the resulting data was used as input to train seven different multivariate classification techniques. The trained classifiers were subsequently implemented and evaluated with respect to their capabilities to correctly predict the classes of unknown fuel items. The best results concerning successful discrimination of UOX and MOX-fuel were acquired when using non-linear classification techniques, such as the k nearest neighbors method and the Gaussian kernel support vector machine. For fuel with cooling times up to 20 years, when it is considered that gamma-rays from the isotope 134Cs can still be efficiently measured, success rates of 100% were obtained. A sensitivity analysis indicated that these methods were also robust.

  1. Non destructive burn up determination of IEA-R1 reactor fuel elements by gamma-ray spectrometry using a Ge(Li) detector

    International Nuclear Information System (INIS)

    Madi Filho, T.

    1982-01-01

    A non destructive determination of burn up of low (IEA-14) and high (IEA-80) activity fuel elements used in the IEA-R1 pool reactor was made from the measured distribution of the Cs-137 gamma-ray activity in these elements. For both series of measurements a 73,7 c.c. Ge(Li) detector was used in 'well collimated' geometry. Where as IEA-14, removed from the reactor some 20 years, showed a gamma-ray spectrum essentially due to Cs-137, IEA-80, with a cooling time of 5 years, showed a more complex spectrum due to the greater number of fission products remaining. The S.I out-of-pool assembly was calibrated using Cs-137 and Co-60 point and Ag-110m plane sources. These measurements provided the necessary constants used to calculate fuel burn-up from measured relative activity distributions of fuel elements. Detailed fuel plate transmission measurements made with the Cs-137 source showed the plates to be highly homogeneous. High activity fuel elements were measured in the S.II in-pool assembly in which the detector was locate on the moveable pool bridge and the test element was positioned immediately below the detector 2.17m below the pool surface. Measurements made in the S.II assembly were normalised with respect to the measured activity of the IEA-14 element. The measured burn up of the IEA-14 and IEA-80 elements obtained in this work is 3.22.10 - 3 gms and 24.44gms. These values may be compared with respective values of 2.63.10 - 3 gms and 61.11gms given by 'total reactor energy/flux distribution' calculations. Calculated errors for the U-235 burn up are 7.4% (IEA-14) and 10.1% (IEA-80). A detailed evaluation of the errors associated with both sets of measurements is given. (Author) [pt

  2. Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

    International Nuclear Information System (INIS)

    Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

    2001-01-01

    The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

  3. Radial power density distribution of MOX fuel rods in the IFA-651

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Ho; Koo, Yang Hyun; Joo, Hyung Kook; Cheon, Jin Sik; Oh, Je Yong; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    Two MOX fuel rods, which were fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with Korea Atomic Energy Research Institute, have been irradiated in the HBWR from June, 2000 in the framework of OECD-HRP together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is basic in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR{sub H}BWR that calculates radial power density distribution for three MOX fuel rods has been developed based on neutron physics results and DEPRESS program. The developed subroutine FACTOR{sub H}BWR gives good agreement with the physics calculation except slight under-prediction at the outer part of the pellet above the burnup of 20 MWd/kgHM. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. 24 figs., 4 tabs. (Author)

  4. Radial power density distribution of MOX fuel rods in the HBWR

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Joo, Hyung Kook; Lee, Byung Ho; Sohn, Dong Seong

    1999-07-01

    Two MOX fuel rods, which ar being fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with the Korea Atomic Energy Research Institute (KAERI), are going to be irradiated in the HBWR (Halden Boiling Water Reactor) from the beginning of 2000 in the framework of OECD Halden Reactor Programme (HRP) together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is a basic property in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR H BWR that calculates radial power density distribution for three MOX fuel rods have been developed subroutine FACTOR H BWR gives good agreement with the physics calculation except slight underprediction in the central part and a little overprediction at the outer part of the pellet. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. (author). 5 refs., 3 tabs., 24 figs

  5. Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Cummins, Dustin Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hollis, Kendall Jon [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Brown, Donald William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-10-18

    This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffraction data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to

  6. Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

    International Nuclear Information System (INIS)

    Cummins, Dustin Ray; Vogel, Sven C.; Hollis, Kendall Jon; Brown, Donald William; Dombrowski, David E.

    2016-01-01

    This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffraction data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to

  7. Thermal and in-pile densification of MOX fuels: Some recent results

    International Nuclear Information System (INIS)

    Caillot, L.; Malgouyres, P.P.; Souchon, F.; Gotta, M.J.; Warin, D.; Chotard, A.; Couty, J.C.

    1997-01-01

    In-pile densification of PWR fuels is one of the main phenomena which determine the evolution of the pellet-clad gap during the first stage of the irradiation, and thus has consequences onto the thermo-mechanical behaviours of fuel rods. It can be predicted using the results of resintering tests and appropriate correlations. In this context, CEA, FRAMATOME and EDF have undertaken a joint research programme aiming to characterize the densification of MOX fuels. Different fuels were prepared by the MIMAS process using different UO 2 powders as matrix. After a detailed characterization, fuel pellets were submitted to isothermal resintering tests and analytical irradiations. Correlations between in-pile and thermal densification were established. This paper presents the results obtained with two types of MOX fuel: one fabricated wit the AUC UO 2 powder (ammonium uranyl carbonate conversion process) and another one fabricated with the SFEROX powder (peroxide conversion process). 8 refs, 8 figs

  8. Application of wavelet scaling function expansion continuous-energy resonance calculation method to MOX fuel problem

    International Nuclear Information System (INIS)

    Yang, W.; Wu, H.; Cao, L.

    2012-01-01

    More and more MOX fuels are used in all over the world in the past several decades. Compared with UO 2 fuel, it contains some new features. For example, the neutron spectrum is harder and more resonance interference effects within the resonance energy range are introduced because of more resonant nuclides contained in the MOX fuel. In this paper, the wavelets scaling function expansion method is applied to study the resonance behavior of plutonium isotopes within MOX fuel. Wavelets scaling function expansion continuous-energy self-shielding method is developed recently. It has been validated and verified by comparison to Monte Carlo calculations. In this method, the continuous-energy cross-sections are utilized within resonance energy, which means that it's capable to solve problems with serious resonance interference effects without iteration calculations. Therefore, this method adapts to treat the MOX fuel resonance calculation problem natively. Furthermore, plutonium isotopes have fierce oscillations of total cross-section within thermal energy range, especially for 240 Pu and 242 Pu. To take thermal resonance effect of plutonium isotopes into consideration the wavelet scaling function expansion continuous-energy resonance calculation code WAVERESON is enhanced by applying the free gas scattering kernel to obtain the continuous-energy scattering source within thermal energy range (2.1 eV to 4.0 eV) contrasting against the resonance energy range in which the elastic scattering kernel is utilized. Finally, all of the calculation results of WAVERESON are compared with MCNP calculation. (authors)

  9. Evaluation of fuel cycle scenarios on MOX fuel recycling in PWRs and SFRs

    Energy Technology Data Exchange (ETDEWEB)

    Carlier, B.; Caron-Charles, M.; Van Den Durpel, L. [AREVA, 1 place Jean Millier, Paris La Defense (France); Senentz, G. [AREVA, 33 rue La Lafayette, 75009 Paris (France); Serpantie, J.P. [AREVA, 10 rue Juliette Recamier, Lyon (France)

    2013-07-01

    Prospects on advanced fuel cycle scenario are considered for achieving a progressive integration of Sodium Fast Reactor (SFR) technology within the current French Pressurized Water Reactor (PWR) nuclear fleet, in a view to benefit from fissile material multi-recycling capability. A step by step process is envisioned, and emphasis is put on its potential implementation through the nuclear mass inventory calculations with the COSAC code. The overall time scale is not optimized. The first step, already implemented in several countries, the plutonium coming from the reprocessing of used Light Water Reactor (LWR) fuels is recycled into a small number of LWRs. The second step is the progressive introduction of the first SFRs, in parallel with the continuation of step 1. This second step lets to prepare the optimized multi recycling of MOX fuel which is considered in step 3. Step 3 is characterized by the introduction of a greater number of SFR and MOX management between EPR reactors and SFRs. In the final step 4, all the fleet is formed with SFRs. This study assesses the viability of each step of the overall scenario. The switch from one step to the other one could result from different constrains related to issues such as resources, waste, experience feedback, public acceptance, country policy, etc.

  10. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    International Nuclear Information System (INIS)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-01-01

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

  11. Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

    International Nuclear Information System (INIS)

    Kudinov, K.G.; Tretyakov, A.A.; Sorokin, Y.P.; Bondin, V.V.; Manakova, L.F.; Jardine, L.J.

    2001-01-01

    In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is incineration

  12. gamma-ray spectra measurements for long cooled MOX spent fuels

    International Nuclear Information System (INIS)

    Murakami, Kiyonobu; Kobayashi, Iwao

    1993-09-01

    Gamma-ray spectra of spent fuels have important informations in the estimation of burnup rate, concentration of fission products, cooling time and etc. which are required in the fuel loading control of reactors and special nuclear materials accountancy from the view point of safe guard. Although, some available data are given about uranium dioxide fuels, few data are given about uranium and plutonium dioxide mixtures (MOX fuels). Especially, there is few data about MOX fuels which are irradiated in thermal reactors and cooled more than ten years. Gamma-ray spectra are measured for PuO 2 -UO 2 fuel rods (IFA-159, IFA-160) which are irradiated at HBWR in Norway up to 9,420 and 5,340MWd/t respectively. Gamma-ray spectra had been measured about the two fuels ten years ago at the spent fuel pond of Japan Demonstration Reactor (JPDR). The objectives of this measurement is to know how decayed the gamma-ray spectra in these ten years and some fission products are there which are effective to estimate burnup rate of spent MOX fuels. (author)

  13. Preliminary analysis of in-reactor behavior of three MOX fuel rods in the halden reactor

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong; Joo, Hyung Kook

    1999-09-01

    Preliminary analysis of in-reactor thermal performance for three MOX fuel rods that are going to be irradiated in the Halden reactor from the first quarter of the year 2000 have been conducted by using the computer code COSMOS. Using the assumption that microstructure of MOX fuel fabricated by SBR and dry milling method is the same, parametric studies have been carried out considering four kinds of uncertainties, which are thermal conductivity, linear power, manufacturing parameters, and model constant, to investigate the effect of each of uncertainty on in-reactor behavior. It is found that the uncertainty of model constants for FGR has a greatest impact of the all because the amount of gas released to the gap is one of the parameters that dominantly affects the gap conductance. The parametric analysis shows that, tn the case of MOX-1, calculational results vary widely depending on the choice of model constants for FGR. Therefore, the model constants for FGR for the present test need to be established through the measured fuel centerline temperature, rod internal pressure, stack length if any, and finally thermal conductivity derived from measured data during irradiation. On the other hand, the difference in thermal performance of MOX-3 resulting from the choice of FGR model constants is not so large as that for MOX-1. This might arise, since the temperature of the MOX-3 is high, the capacity of grain boundaries to retain gas atoms is not sufficient enough to accommodate the large amount of gas atoms reaching the grain boundaries through diffusion. (Author). 20 refs., 7 tabs., 47 figs

  14. Preliminary analysis of in-reactor behavior of three MOX fuel rods in the halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong; Joo, Hyung Kook

    1999-09-01

    Preliminary analysis of in-reactor thermal performance for three MOX fuel rods that are going to be irradiated in the Halden reactor from the first quarter of the year 2000 have been conducted by using the computer code COSMOS. Using the assumption that microstructure of MOX fuel fabricated by SBR and dry milling method is the same, parametric studies have been carried out considering four kinds of uncertainties, which are thermal conductivity, linear power, manufacturing parameters, and model constant, to investigate the effect of each of uncertainty on in-reactor behavior. It is found that the uncertainty of model constants for FGR has a greatest impact of the all because the amount of gas released to the gap is one of the parameters that dominantlyaffects the gap conductance. The parametric analysis shows that, tn the case of MOX-1, calculational results vary widely depending on the choice of model constants for FGR. Therefore, the model constants for FGR for the present test need to be established through the measured fuel centerline temperature, rod internal pressure, stack length if any, and finally thermal conductivity derived from measured data during irradiation. On the other hand, the difference in thermal performance of MOX-3 resulting from the choice of FGR model constants is not so large as that for MOX-1. This might arise, since the temperature of the MOX-3 is high, the capacity of grain boundaries to retain gas atoms is not sufficient enough to accommodate the large amount of gas atoms reaching the grain boundaries through diffusion. (Author). 20 refs., 7 tabs., 47 figs.

  15. Determining plutonium mass in spent fuel with non-destructive assay techniques - NGSU research overview and update on 6 NDA techniques

    Energy Technology Data Exchange (ETDEWEB)

    Tobin, Stephen J [Los Alamos National Laboratory; Conlin, Jeremy L [Los Alamos National Laboratory; Evans, Louise G [Los Alamos National Laboratory; Hu, Jianwei [Los Alamos National Laboratory; Blanc, Pauline C [Los Alamos National Laboratory; Lafleur, Adrienne M [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Schear, Melissa A [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Fensin, Michael L [Los Alamos National Laboratory; Freeman, Corey R [Los Alamos National Laboratory; Koehler, William E [Los Alamos National Laboratory; Mozin, V [Los Alamos National Laboratory; Sandoval, N P [Los Alamos National Laboratory; Lee, T H [KAERI; Cambell, L W [PNNL; Cheatham, J R [ORNL; Gesh, C J [PNNL; Hunt, A [IDAHO STATE UNIV; Ludewigt, B A [LBNL; Smith, L E [PNNL; Sterbentz, J [INL

    2010-09-15

    This poster is one of two complementary posters. The Next Generation Safeguards Initiative (NGSI) of the U.S. DOE has initiated a multi-lab/university collaboration to quantify the plutonium (Pu) mass in, and detect the diversion of pins from, spent nuclear fuel assemblies with non-destructive assay (NDA). This research effort has the goal of quantifying the capability of 14 NDA techniques as well as training a future generation of safeguards practitioners. By November of 2010, we will be 1.5 years into the first phase (2.5 years) of work. This first phase involves primarily Monte Carlo modelling while the second phase (also 2.5 years) will focus on experimental work. The goal of phase one is to quantify the detection capability of the various techniques for the benefit of safeguard technology developers, regulators, and policy makers as well as to determine what integrated techniques merit experimental work, We are considering a wide range of possible technologies since our research horizon is longer term than the focus of most regulator bodies. The capability of all of the NDA techniques will be determined for a library of 64 17 x 17 PWR assemblies [burnups (15, 30, 45, 60 GWd/tU), initial enrichments (2, 3, 4, 5%) and cooling times (1, 5, 20, 80 years)]. The burnup and cooling time were simulated with each fuel pin being comprised of four radial regions. In this paper an overview of the purpose will be given as well as a technical update on the following 6 neutron techniques: {sup 252}Cf Interrogation with Prompt Neutron Detection, Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Passive Neutron Albedo Reactivity, Self-Integration Neutron Resonance Densitometry. The technical update will quantify the anticipated performance of each technique for the 64 assemblies of the spent fuel library.

  16. Determining plutonium mass in spent fuel with non-destructive assay techniques - NGSI research overview and update on 6 NDA techniques

    International Nuclear Information System (INIS)

    Tobin, Stephen J.; Conlin, Jeremy L.; Evans, Louise G.; Hu, Kianwei; Blanc, P.C.; Lafleur, Am; Menlove, H.O.; Schear, M.A.; Swinhoe, M.T.; Croft, S.; Fensin, M.L.; Freeman, C.R.; Koehler, W.E.; Mozin, V.; Sandoval, N.P.; Lee, T.H.; Cambell, L.W.; Cheatham, J.R.; Gesh, C.J.; Hunt, A.; Ludewigt, B.A.; Smith, L.E.; Sterbentz, J.

    2010-01-01

    This poster is one of two complementary posters. The Next Generation Safeguards Initiative (NGSI) of the U.S. DOE has initiated a multi-lab/university collaboration to quantify the plutonium (Pu) mass in, and detect the diversion of pins from, spent nuclear fuel assemblies with non-destructive assay (NDA). This research effort has the goal of quantifying the capability of 14 NDA techniques as well as training a future generation of safeguards practitioners. By November of 2010, we will be 1.5 years into the first phase (2.5 years) of work. This first phase involves primarily Monte Carlo modelling while the second phase (also 2.5 years) will focus on experimental work. The goal of phase one is to quantify the detection capability of the various techniques for the benefit of safeguard technology developers, regulators, and policy makers as well as to determine what integrated techniques merit experimental work, We are considering a wide range of possible technologies since our research horizon is longer term than the focus of most regulator bodies. The capability of all of the NDA techniques will be determined for a library of 64 17 x 17 PWR assemblies (burnups (15, 30, 45, 60 GWd/tU), initial enrichments (2, 3, 4, 5%) and cooling times (1, 5, 20, 80 years)). The burnup and cooling time were simulated with each fuel pin being comprised of four radial regions. In this paper an overview of the purpose will be given as well as a technical update on the following 6 neutron techniques: 252 Cf Interrogation with Prompt Neutron Detection, Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Passive Neutron Albedo Reactivity, Self-Integration Neutron Resonance Densitometry. The technical update will quantify the anticipated performance of each technique for the 64 assemblies of the spent fuel library.

  17. Innovation in Non Destructive Testing

    NARCIS (Netherlands)

    Wassink, C.H.P.

    2012-01-01

    In many established companies the pace of innovation is low. The Non-Destructive Testing sector is an example of a sector where the pace of innovation is very slow. Non-Destructive Testing (NDT) refers to the set of non-invasive activities used to determine the condition of objects or installations

  18. A PCI failure in an experimental MOX fuel rod and its sensitivity analysis

    International Nuclear Information System (INIS)

    Marino, A.C.

    2000-01-01

    Within our interest in studying MOX fuel performance, the irradiation of the first Argentine prototypes of PHWR MOX fuels began in 1986 with six rods fabricated at the α Facility (CNEA, Argentina). These experiences were made in the HFR-Petten reactor, Holland. The goal of this experience was to study the fuel behaviour with respect to PMCI-SCC. An experiment for extended burnup was performed with the last two MOX rods. During the experiment the final test ramp was interrupted due to a failure in the rod. The post-irradiation examinations indicated that PCI-SCC was a mechanism likely to produce the failure. At the Argentine Atomic Energy Commission (CNEA) the BACO code was developed for the simulation of a fuel rod thermo-mechanical behaviour under stationary and transient power situations. BACO includes a probability analysis within its structure. In BACO the criterion for safe operation of the fuel is based on the maximum hoop stress being below a critical value at the cladding inner surface; this is related to susceptibility to stress corrosion cracking (SCC). The parameters of the MOX irradiation, the preparation of the experiments and post-irradiation analysis were sustained by the BACO code predictions. We present in this paper an overview of the different experiences performed with the MOX fuel rods and the main findings of the post-irradiation examinations. A BACO code description, a wide set of examples which sustain the BACO code validation, and a special calculation for BU15 experiment attained using the BACO code, including a probabilistic analysis of the influence of rod parameters on performance, are included. (author)

  19. Image analysis and 2D/3D modeling of the MOX fuel microstructure

    International Nuclear Information System (INIS)

    Oudinet, Ghislain

    2003-01-01

    The microstructure of the MOX fuel, made with UO_2 and PuO_2, determines his 'in pile' behavior. The french companies CEA and COGEMA are highly interested in its description by image analysis, which is the object of the present work. The segmentation algorithms described here use pictures issued from a microprobe and a SEM, to analyse the plutonium and porosity distribution in the fuel pellets. They are innovating, automated and robust enough to be used with a small data set. They have been successfully tested on different fuels, before and after irradiation. Three-dimensional informations have been computed with a genetic algorithm. The obtained 3D object size distributions allowed the modeling of many different industrial and research fuels. 3D reconstruction is accurate and stable, and provides a basis for different studies among which the study of the MOX fuel 'in pile' behavior. (author)

  20. The development of a commercial MOX fuel manufacturing capability in the U.K

    International Nuclear Information System (INIS)

    Macphee, D.S.; Young, M.P.

    1995-01-01

    BNFL is implementing a strategy to establish a commercial MOX manufacturing capability within the UK. The design and provision of the fabrication plants is incorporating the considerable experience within the Company of MOX technology, fuel fabrication and nuclear plant design. The first phase of the strategy is complete with the successful operation of the Demonstration Facility. The development programmes supporting the increased scale of operation for a commercial scale facility are substantially complete. Design and construction of a 120t HM/year plant is well advanced supported by a substantial in-house design and project management team. (author)

  1. Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vapirev, E; Jordanov, T; Khristoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1996-12-31

    The possibility of using highly enriched uranium available from military inventories for production of mixed oxide fuel (MOX) has been proposed. The fuel is based on U-235 dioxide as fissile isotope and Th-232 dioxide as a non-fissile isotope. It is shown that although the fuel conversion coefficient to U-233 is expected to be less than 1, the proposed fuel has several important advantages resulting in cost reduction of the nuclear fuel cycle. The expected properties of MOX fuel (cross-sections, generated chains, delayed neutrons) are estimated. Due to fuel generation the initial enrichment is expected to be 1% less for production of the same energy. In contrast to traditional fuel no long living actinides are generated which reduces the disposal and reprocessing cost. 7 refs.

  2. A fission gas release model for MOX fuel and its verification

    International Nuclear Information System (INIS)

    Koo, Y.H.; Sohn, D.S.; Strijov, P.

    2000-01-01

    A fission gas release model for MOX fuel has been developed based on a model for UO 2 fuel. Using the concept of equivalent cell, the model considers the uneven distribution of Pu within the fuel matrix and a number of Pu-rich particles that could lead to a non-uniform fission rate and fission gas distribution across the fuel pellet. The model has been incorporated into a code, COSMOS, and some parametric studies were made to analyze the effect of the size and Pu content of Pu-rich agglomerates. The model was then applied to the experimental data obtained from the FIGARO program, which consisted of the base irradiation of MOX fuels in the BEZNAU-1 PWR and the subsequent irradiation of four refabricated fuel segments in the Halden reactor. The calculated gas releases show good agreement with the measured ones. In addition, the present analysis indicates that the microstructure of the MOX fuel used in the FIGARO program is such that it has produced little difference in terms of gas release compared with UO 2 fuel. (author)

  3. Hanford MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    O'Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site (SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. Hanford has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 1 facility. In all, a total of three LA MOX fuel fabrication options were identified by Hanford that could accommodate the program. In every case, only minor modification would be required to ready any of the facilities to accept the equipment necessary to accomplish the LA program

  4. Hanford MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site (SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. Hanford has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 1 facility. In all, a total of three LA MOX fuel fabrication options were identified by Hanford that could accommodate the program. In every case, only minor modification would be required to ready any of the facilities to accept the equipment necessary to accomplish the LA program.

  5. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO 2 powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO 2 powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule

  6. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule.

  7. Parametric study on co-precipitation of U/Th in MOX fuel of AHWR

    International Nuclear Information System (INIS)

    Tiwari, S.K.; Swaroopa Lakshmi, Y.; Nath, Baidurjya; Setty, D.S.; Kalyana Krishnan, G.; Saibaba, N.

    2015-01-01

    During manufacturing of Mixed Oxide Fuel (MOX) pellets for Advance Heavy Water Reactor (AHWR-LEU), around 30% rejected MOX pellets are generated in every cycle. These rejected MOX pellets are dissolved in nitric acid for recovery of U/Th. The recovered U/Th is recycled for production of MOX pellets. MOX pellets of varying compositions are used in AHWR fuel. Dissolution of MOX pellets in nitric acid is a challenging task because of its low surface area and longer dissolution times. High normal nitric acid is used in order to increase rate of dissolution, which in turn results in generation of high free acidity solution which influences the precipitation characteristics of Uranium (VI) by oxalic acid. Oxalic acid precipitation helps in generation of nitric acid which can be used for dissolution there by effectively facilitating nil effluent generation. Precipitation by oxalic acid unlike ammonia has advantage of zero liquid effluent discharge by complete recycle of oxalate filtrate to dissolution section. In the present work, the effect of various parameters like free acidity, residence time, concentration of oxalic acid, initial concentration of uranium and thorium etc. on the precipitation of U(VI) and Th(IV) in nitrate media by oxalic acid was carried out. The precipitated powder was subjected to various morphological evaluations like particle size etc. Study of various parameters on the co-precipitation of uranium and thorium by oxalic acid was carried out. It was observed that complete precipitation (> 99.9%) of thorium as oxalate does not depend on free acidity range (1- 6 N). Excess oxalic acid is not required for complete precipitation of thorium oxalate. The precipitation of uranyl oxalate varies with initial free acidity of solution. Uranyl oxalate precipitation does not take place at and above 5 N of free acidity

  8. Implement of MOX fuel assemblies in the design of the fuel reload for a BWR; Implemento de ensambles de combustible MOX en el diseno de la recarga de combustible para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Enriquez C, P.; Ramirez S, J. R.; Alonso V, G.; Palacios H, J. C., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    At the present time the use of mixed oxides as nuclear fuel is a technology that has been implemented in mixed reloads of fuel for light water reactors. Due to the plutonium production in power reactors, is necessary to realize a study that presents the plutonium use like nuclear fuel. In this work a study is presented that has been carried out on the design of a fuel assembly with MOX to be proposed in the supply of a fuel reload. The fissile relationship of uranium to plutonium is presented for the design of the MOX assembly starting from plutonium recovered in the reprocessing of spent fuel and the comparison of the behavior of the infinite multiplication factor is presented and of the local power peak factor, parameters of great importance in the fuel assemblies design. The study object is a fuel assembly 10 x 10 GNF2 type for a boiling water reactor. The design of the fuel reload pattern giving fuel assemblies with MOX, so the comparison of the behavior of the stop margin for a fuel reload with UO{sub 2} and a mixed reload, implementing 12 and 16 fuel assemblies with MOX are presented. The results show that the implement of fuel assemblies with MOX in a BWR is possible, but this type of fuels creates new problems that are necessary to study with more detail. In the development of this work the calculus tools were the codes: INTREPIN-3, CASMO-4, CMSLINK and SIMULATE-3. (Author)

  9. Development and validation of the ENIGMA code for MOX fuel performance modelling

    International Nuclear Information System (INIS)

    Palmer, I.; Rossiter, G.; White, R.J.

    2000-01-01

    The ENIGMA fuel performance code has been under development in the UK since the mid-1980s with contributions made by both the fuel vendor (BNFL) and the utility (British Energy). In recent years it has become the principal code for UO 2 fuel licensing for both PWR and AGR reactor systems in the UK and has also been used by BNFL in support of overseas UO 2 and MOX fuel business. A significant new programme of work has recently been initiated by BNFL to further develop the code specifically for MOX fuel application. Model development is proceeding hand in hand with a major programme of MOX fuel testing and PIE studies, with the objective of producing a fuel modelling code suitable for mechanistic analysis, as well as for licensing applications. This paper gives an overview of the model developments being undertaken and of the experimental data being used to underpin and to validate the code. The paper provides a summary of the code development programme together with specific examples of new models produced. (author)

  10. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses - Revision 1

    International Nuclear Information System (INIS)

    Hermann, O.W.

    2000-01-01

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotopes) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data, usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, was considered to be of sufficient quality for depletion code validation

  11. Non-linear behaviour of multi-phase MOX fuels: a micro-mechanical approach

    International Nuclear Information System (INIS)

    Rousette, S.; Gatt, J.M.; Michel, J.C.

    2005-01-01

    The modelling of mechanical pellet-clad interaction requires knowledge of the thermo-mechanical behaviour of nuclear fuels. Some nuclear fuels such as MOX are composed of several phases. The mechanical properties of these phases, which are elasto-visco-plastic in-pile, are changing in-pile. The objective is to formulate a mechanical behaviour law taking all the physical phenomena into account in the different phases, which can easily be introduced into a fuel rod modelling code. Consequently, Non-uniform Transformation Field Analysis (NTFA) is used on the one hand, to correctly capture the heterogeneity of the anelastic strain in the different phases and, on the other hand, to provide a simple overall constitutive law for computational codes. This method is a good way to describe the behaviour of MOX fuel. Transformation Field Analysis (TFA), which corresponds to piecewise uniform transformation fields, is used to perform a sensitivity study. (authors)

  12. Highlights on R and D work related to the achievement of high burnup with MOX fuel in commercial reactors

    International Nuclear Information System (INIS)

    Lippens, M.; Maldague, Th.; Basselier, J.; Boulanger, D.; Mertens, L.

    2000-01-01

    Part of the R and D work made at BELGONUCLEAIRE in the field of high burnup achievement with MOX fuel in commercial LWRs is made through lnternational Programmes. Special attention is given to the evolution with burnup of fuel neutronic characteristics and of in-reactor rod thermal-mechanical behaviour. Pu burning in MOX is characterized essentially by a drop of Pu 239 content. The other Pu isotopes have an almost unchanged concentration, due to internal breeding. The reactivity drop of MOX versus burnup is consequently much less pronounced than in UO 2 fuel. Concentration of minor actinides Am and Cm becomes significant with burnup increase. These nuclides start to play a role on total reactivity and in the helium production. The thermal-mechanical behaviour of MOX fuel rod is very similar to that of UO 2 . Some specificities are noticed. The better PCI resistance recognized to MOX fuel has recently been confirmed. Three PWR MOX segments pm-irradiated up to 58 GWd/tM were ramped at 100 W/cm.min respectively to 430-450-500 W/cm followed by a hold time of 24 hours. No segment failed. MOX and UO 2 fuels have different reactivities and operate thus at different powers. Moreover, radial distribution of power in MOX pellet is less depressed at high burnup than in UO 2 , leading to higher fuel central temperature for a same rating. The thermal conductivity of MOX fuel decreases with Pu content, typically 4% for 10% Pu. The combination of these three elements (power level, power profile, and conductivity) lead to larger FGR at high burnup compared to UO 2 . Helium production remains low compared to fission gas production (ratio < 0.2). As faster diffusing element, the helium fractional release is much higher than that of fission gas, leading to rod pressure increase comparable to the one resulting from fission gas. (author)

  13. Advanced chemical quality control techniques for use in the manufacture of (U-Pu) MOX fuels

    International Nuclear Information System (INIS)

    Panakkal, J.P.; Prakash, Amrit

    2010-01-01

    Analytical chemistry plays a very important role for nuclear fuel cycle activities be it fuel fabrication, waste management or reprocessing. Nuclear fuels are selected based on the type of reactor. The nuclear fuel has to conform to various stringent chemical specifications like B, rare earths, H, O/M heavy metal content etc. Selection of technique is very important to determine the true specification. This is important particularly when the analyses has to be performed inside leak tight enclosure. The present paper describes the details of the advanced techniques being developed and used in the manufacture of (U,Pu) MOX fuels. (author)

  14. Non-destructive inservice inspections

    International Nuclear Information System (INIS)

    Kauppinen, P.; Sarkimo, M.; Lahdenperae, K.

    1998-01-01

    In order to assess the possible damages occurring in the components and structures of operating nuclear power plants during service the main components and structures are periodically inspected by non-destructive testing techniques. The reliability of non-destructive testing techniques applied in these inservice inspections is of major importance because the decisions concerning the needs for repair of components are mainly based on the results of inspections. One of the targets of this research program has been to improve the reliability of non-destructive testing. This has been addressed in the sub-projects which are briefly summarised here. (author)

  15. Performance of cladding on MOX fuel with low 240Pu/239Pu ratio

    International Nuclear Information System (INIS)

    McCoy, K.; Blanpain, P.; Morris, R.

    2015-01-01

    The U.S. Department of Energy has decided to dispose of a portion of its surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. As part of fuel qualification, four lead assemblies were manufactured and irradiated to a maximum fuel rod average burnup of 47.3 MWd/kg heavy metal. This was the world's first commercial irradiation of MOX fuel with a 240 Pu/ 239 Pu ratio less than 0.10. Five fuel rods with varying burnups and plutonium contents were selected from one of the assemblies and shipped to Oak Ridge National Laboratory for hot cell examination. This paper discusses the results of those examinations with emphasis on cladding performance. Exams relevant to the cladding included visual and eddy current exams, profilometry, microscopy, hydrogen analysis, gallium analysis, and mechanical testing. There was no discernible effect of the type of MOX fuel on the performance of the cladding. (authors)

  16. Beginning-of-life gap closure behaviour of experimental PFBR MOX fuel pin

    International Nuclear Information System (INIS)

    Jayaraj, V.V.; Padalakshmi, M.; Ojha, B.K.; Padma Prabu, C.; Saravanan, T.; Venkiteswaran, C.N.; Philip, John; Muralidharan, N.G.; Joseph, Jojo; Kasiviswanathan, K.V.; Jayakumar, T.

    2011-01-01

    Mixed oxide fuel with 22 % and 29% plutonium is chosen as the fuel for PFBR for the two fissile zones. Due to the fabrication tolerances in the pellet diameter, fuel has to be preconditioned at a lower linear power for a brief period before raising the power to the rated value of 450 W/cm. PIE was done on an experimental MOX fuel pin irradiated in FBTR for 13 days at a linear power of 400 W/cm for gap closure studies with the objective of optimising the duration of pre-conditioning before raising the power to the design value of 450 W/cm. X-radiography and remote metallography was done on the fuel pin to estimate the axial fuel column elongation and fuel-clad gap. Remote metallography of the fuel pin cross-sections at five axial locations of the fuel column and the subsequent fuel-clad gap measurement has indicated that the average radial gap has reduced from the pre-irradiation value of 75-110 microns to around 12-13 microns along the entire length of the fuel column. This paper will describe the details of examinations and results of the PIE carried out on the MOX fuel pin. (author)

  17. Application of non-destructive methods for qualification of the U3O8-Al and U3Si2-Al dispersion fuels in the IEA-R1 Reactor

    International Nuclear Information System (INIS)

    Silva, Jose Eduardo Rosa da

    2011-01-01

    IPEN/CNEN-SP manufactures fuels to be used in its nuclear research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil doesn't have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds, internationally tested and qualified to be used in research reactors, and has gotten experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans to increase the uranium density of these fuels. The objective of this thesis work was to study and to propose a set of non-destructive methods to qualify the dispersions fuels U 3 O 8 -Al e U 3 Si 2 -Al with high uranium density produced at IPEN/CNEN-SP. For that, the irradiation resources in the IEA-R1, and the application of non-destructive methods in the reactor pool available in the Institution were considered. The proposal is to specify, manufacture and irradiate fuel mini plates in IEA-R1 at the maximum densities, qualified internationally, and to monitor their general conditions during the period of irradiation, using non-destructive methods in the reactor pool. In addition to the non-destructive visual inspection and sipping methods, already used at the Institution, the infrastructure for dimensional sub-aquatic testing to evaluate the swelling of irradiated fuel mini plates was completed. The analyses of the results will provide means to assess and decide whether or not to continue with the irradiation of mini plates, until the desired burnup for the irradiation tests at IEA-R1 are reached. (author)

  18. Overall models and experimental database for UO2 and MOX fuel increasing performance

    International Nuclear Information System (INIS)

    Bernard, L.C.; Blanpain, P.

    2001-01-01

    COPERNIC is an advanced fuel rod performance code developed by Framatome. It is based on the TRANSURANUS code that contains a clear and flexible architecture, and offers many modeling possibilities. The main objectives of COPERNIC are to accurately predict steady-state and transient fuel operations at high burnups and to incorporate advanced materials such as the Framatome M5-alloy cladding. An extensive development program was undertaken to benchmark the code to very high burnups and to new M5-alloy cladding data. New models were developed for the M5-alloy cladding and the COPERNIC thermal models were upgraded and improved to extend the predictions to burnups over 100 GWd/tM. Since key phenomena, like fission gas release, are strongly temperature dependent, many other models were upgraded also. The COPERNIC qualification range extends to 67, 55, 53 GWd/tM respectively for UO 2 , UO 2 -Gd 2 O 3 , and MOX fuels with Zircaloy-4 claddings. The range extends to 63 GWd/tM with UO 2 fuel and the advanced M5-alloy cladding. The paper focuses on thermal and fission gas release models, and on MOX fuel modeling. The COPERNIC thermal model consists of several submodels: gap conductance, gap closure, fuel thermal conductivity, radial power profile, and fuel rim. The fuel thermal conductivity and the gap closure models, in particular, have been significantly improved. The model was benchmarked with 3400 fuel centerline temperature data from many French and international programs. There are no measured to predicted statistical biases with respect to linear heat generation rate or burnup. The overall quality of the model is state-of-the-art as the model uncertainty is below 10 %. The fission gas release takes into account athermal and thermally activated mechanisms. The model was adapted to MOX and Gadolinia fuels. For the heterogeneous MOX MIMAS fuels, an effective burnup is used for the incubation threshold. For gadolinia fuels, a scaled temperature effect is used. The

  19. Burn-up credit applications for UO2 and MOX fuel assemblies in AREVA/COGEMA

    International Nuclear Information System (INIS)

    Toubon, H.; Riffard, C.; Batifol, M.; Pelletier, S.

    2003-01-01

    For the last seven years, AREVA/COGEMA has been implementing the second phase of its burn-up credit program (the incorporation of fission products). Since the early nineties, major actinides have been taken into account in criticality analyses first for reprocessing applications, then for transport and storage of fuel assemblies Next year (2004) COGEMA will take into account the six main fission products (Rh103, Cs133, Nd143, Sm149, Sm152 and Gd155) that make up 50% of the anti-reactivity of all fission products. The experimental program will soon be finished. The new burn-up credit methodology is in progress. After a brief overview of BUC R and D program and COGEMA's application of the BUC, this paper will focus on the new burn-up measurement for UO2 and MOX fuel assemblies. It details the measurement instrumentation and the measurement experiments on MOX fuels performed at La Hague in January 2003. (author)

  20. Thorium utilization in a small long-life HTR. Part I: Th/U MOX fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands)

    2014-02-15

    Highlights: • We propose thorium MOX (TMOX) fuel blocks for a small block-type HTR. • The TMOX fuel blocks with low-enriched uranium are recommended. • More thorium decreases the reactivity swing of the TMOX fuel blocks. • Thorium reduces the negative temperature coefficient of the TMOX fuel blocks. • Thorium increases the conversion ratio of the TMOX fuel blocks. - Abstract: The U-Battery is a small, long-life and transportable high temperature gas-cooled reactor (HTR). The neutronic features of a typical fuel block with uranium and thorium have been investigated for a application of the U-Battery, by parametrically analyzing the composition and geometric parameters. The type of fuel block is defined as Th/U MOX fuel block because uranium and thorium are assumed to be mixed in each fuel kernel as a form of (Th,U)O{sub 2}. If the initially loaded mass of U-235 is mostly consumed in the early period of the lifetime of Th/U MOX fuel block, low-enriched uranium (LEU) as ignited fuel will not largely reduce the neutronic performance of the Th/U MOX fuel block, compared with high-enriched uranium. The radii of fuel kernels and fuel compacts and packing fraction of TRISO particles determine the atomic ratio of the carbon to heavy metal. When the ratio is smaller than 400, the difference among them due to double heterogeneous effects can be neglected for the Th/U MOX fuel block. In the range between 200 and 400, the reactivity swing of the Th/U MOX fuel block during 10 years is sufficiently small. The magnitude of the negative reactivity temperature coefficients of the Th/U MOX fuel block decreases by 20–45%, which is positive to reduce temperature defect of the Th/U MOX fuel block. The conversion ratio (CR) of the fuel increases from 0.48 (typical CR of the LEU-fueled U-Battery) to 0.78. The larger conversion ratio of the Th/U MOX fuel block reduces the reactivity swing during 10 years for the U-Battery.

  1. The MOX

    International Nuclear Information System (INIS)

    Legay, Christophe

    1997-06-01

    In this report, the author first proposes a presentation of plutonium with a brief history of its discovery and the discovery of other transuranic elements, a presentation of its main characteristics, and a description of its production ways. He also proposes an overview of data regarding world plutonium production and plutonium stock situation. The second part addresses the MOX fuel in relationship with the choice of non proliferation. The author describes the MOX fuel cycle (production, use in reactor, and reprocessing) and outlines the environmental and economic benefits of this fuel, and its interest within the frame of struggle against nuclear proliferation. The third part addresses the present situation and perspectives. He comments the American posture (principles and recent statements), discusses alternatives regarding nuclear wastes, and outlines MOX opportunities by evoking the French case and international perspectives, and the benefits in terms of matching irreversibility and safety

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

    International Nuclear Information System (INIS)

    Ito, Chihiro

    1999-01-01

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

  3. Behavior of irradiated ATR/MOX fuel under reactivity initiated accident conditions (Joint research)

    International Nuclear Information System (INIS)

    Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Uetsuka, Hiroshi

    2000-03-01

    Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20 MWd/kgHM were conducted at the NSRR in JAERI to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Deformation of the fuel rods due to PCMI occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet. (author)

  4. Characterization of un-irradiated MIMAS MOX fuel by Raman spectroscopy and EPMA

    Science.gov (United States)

    Talip, Zeynep; Peuget, Sylvain; Magnin, Magali; Tribet, Magaly; Valot, Christophe; Vauchy, Romain; Jégou, Christophe

    2018-02-01

    In this study, Raman spectroscopy technique was implemented to characterize un-irradiated MIMAS (MIcronized - MASter blend) MOX fuel samples with average 7 wt.% Pu content and different damage levels, 13 years after fabrication, one year after thermal recovery and soon after annealing, respectively. The impacts of local Pu content, deviation from stoichiometry and self-radiation damage on Raman spectrum of the studied MIMAS MOX samples were assessed. MIMAS MOX fuel has three different phases Pu-rich agglomerate, coating phase and uranium matrix. In order to distinguish these phases, Raman results were associated with Pu content measurements performed by Electron Microprobe Analysis. Raman results show that T2g frequency significantly shifts from 445 to 453 cm-1 for Pu contents increasing from 0.2 to 25 wt.%. These data are satisfactorily consistent with the calculations obtained with Gruneisen parameters. It was concluded that the position of the T2g band is mainly controlled by Pu content and self-radiation damage. Deviation from stoichiometry does not have a significant influence on T2g band position. Self-radiation damage leads to a shift of T2g band towards lower frequency (∼1-2 cm-1 for the UO2 matrix of damaged sample). However, this shift is difficult to quantify for the coating phase and Pu agglomerates given the dispersion of high Pu concentrations. In addition, 525 cm-1 band, which was attributed to sub-stoichiometric structural defects, is presented for the first time for the self-radiation damaged MOX sample. Thanks to the different oxidation resistance of each phase, it was shown that laser induced oxidation could be alternatively used to identify the phases. It is demonstrated that micro-Raman spectroscopy is an efficient technique for the characterization of heterogeneous MOX samples, due to its low spatial resolution.

  5. SMOPY, a new NDA tool for safeguards of LEU and MOX spent fuel

    International Nuclear Information System (INIS)

    Lebrun, A.; Merelli, M.; Szabo, J.-L.; Huver, M.; Arenas-Carrasco, J.

    2001-01-01

    Upon IAEA request, the French support program to IAEA Safeguards has developed a new device for control of the irradiated LEU and MOX fuels. The Safeguards Mox Python (SMOPY) is the achievement of a 4 years R and D program supported by CEA and COGEMA in partnership with Eurisys Mesures. The SMOPY system is based on the combination of 2 NDA techniques (passive neutron and room temperature gamma spectrometry) and on line interpretation tools (automatic gamma spectrum interpretation, depletion code EVO). Through the measurement managing software, all this contributes to the fully automatic measurement, interpretation and characterization of any kind of spent fuel. The device is transportable (50 kg, 60 cm) and is composed of four parts: 1. the measurement head with one high efficiency fission chamber and a micro room temperature gamma spectrometric probe; 2. the carrier which carries the measurement head. The carrier bottom fits the racks for accurate positioning and its top fits operator's fuel moving tool; 3. the portable electronic cabinet which includes both neutron and gamma electronic cards; 4. the portable PC which gets inspectors data, controls the measurement, get measured values, interprets them and immediately provides the inspector with worthwhile info for appropriate on the field decisions. Main features of SMOPY are: Discrimination of MOX versus LEU irradiated fuels in any case (conservative case is one cycle MOX versus three cycles LEU after short cooling time); Full characterization of irradiated LEU (burnup, cooling time, Pu amounts ...); Partial Defect Test on LEU fuels. A first version of SMOPY has been tested in industrial condition during summer 2000. This tests shown a need of shielding improvement around the gamma detector. A new version has been build a will be qualified during a new field test and then the system will be ready for routine operation in IAEA and commercial delivery. After giving details about the system itself, this paper

  6. Fuel cycle and waste management. 2. Design of a BWR Core with Over-moderated MOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Francois, J.L.; Del Campo, C. Martin

    2001-01-01

    The use of uranium-plutonium mixed-oxide (MOX) fuel in light water reactors is a current practice in several countries. Generally one-third of the reactor core is loaded with MOX fuel assemblies, and the other two-thirds is loaded with uranium assemblies. Nevertheless, the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this work, the design of a boiling water reactor (BWR) core fully loaded with over-moderated MOX fuel designs was investigated. In previous work, the design of over-moderated BWR MOX fuel assemblies based on a 10 x 10 lattice was presented; these designs improve the neutron spectrum and the plutonium consumption rate, compared with standard MOX assemblies. To increase the moderator-to-fuel ratio (MFR), two approaches were followed. In the first approach, 8 or 12 fuel rods were replaced by water rods in the 10x10 assembly, which increased the MFR from 1.9 to 2.2 and 2.4, respectively. These designs are called MOX-8WR and MOX-12WR, respectively, in this paper. In the second approach, an 11 x 11 lattice with 24 water rods (11 x 11-24WR) was designed, which is a design with a number of active fuel rods (88) very close to the standard MOX assembly (91). The fuel rod diameter is smaller to preserve the assembly dimensions, and in this last case, the MFR is 2.4. The calculations were performed with the CM-PRESTO three-dimensional steady-state simulator. The nuclear data banks were generated with the HELIOS system, and they were processed by TABGEN to produce tables of nuclear cross sections depending on burnup, void, and exposure weighted void (void history), which are used by CM-PRESTO. One base reload pattern was designed for a BWR/5 rated at 1931 MW(thermal), to be used with the different over-moderated assembly designs. The reload pattern has 112 fresh fuel assemblies (FFAs) out of a total of 444 fuel assemblies and was simulated during 20 cycles with the Haling strategy, until an equilibrium cycle of

  7. Advanced PWR Core Design with Siemens High-Plutonium-Content MOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Dieter Porsch; Gerhard Schlosser; Hans-Dieter Berger

    2000-01-01

    The Siemens experience with plutonium recycling dates back to the late 1960s. Over the years, extensive research and development programs were performed for the qualification of mixed-oxide (MOX) technology and design methods. Today's typical reload enrichments for uranium and MOX fuel assemblies and modern core designs have become more demanding with respect to accuracy and reliability of design codes. This paper presents the status of plutonium recycling in operating high-burnup pressurized water reactor (PWR) cores. Based on actual examples, it describes the validation status of the design methods and stresses current and future needs for fuel assembly and core design including those related to the disposition of weapons-grade plutonium

  8. Burning of MOX fuels in LWRs; fuel history effects on thermal properties of hull and end piece wastes and the repository performance

    International Nuclear Information System (INIS)

    Hirano, Fumio; Sato, Seichi; Kozaki, Tamotsu

    2012-01-01

    The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the history of MOX fuels. This history includes the burn-up of UO 2 spent fuels from which the Pu is obtained, the cooling period before reprocessing, the storage period of fresh MOX fuels before being loaded into an LWR, as well as the burn-up of the MOX fuels. The heat generation rates in hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO 2 spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80degC is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 and 70 GWd-MOX needs to be limited to a value of 0.4-1.6, which is significantly lower than 4.0 for 45 GWd-UO 2 . (author)

  9. Non-Destructive Methods for Determining Burn-Up in Nuclear Fuel; Methodes Non Destructives d'Evaluation du Taux de Combustion dans le Combustible Nucleaire; Metody opredeleniya vygoraniya v yadernom toplive bez razrusheniya obraztsa; Metodos No Destructivos para Determinai el Grado de Combustion de los Elementos Combustibles Nucleares

    Energy Technology Data Exchange (ETDEWEB)

    McGonnagle, W. J. [Illinois Institute of Technology, Chicago, IL (United States)

    1966-02-15

    Non-destructive methods for quantitative measurement of burn-up in nuclear reactor fuel elements are useful and desirable. The ideal method for fuel assay would be one that requires no special information about the neutron spectra, radiation history, or cooling time. The irradiated fuel element contains a record of the fuel burn-up. This record is in the form of radioactive and stable isotopes resulting from the fission process. Unfortunately, in the non-destructive as well as the destructive fuel assay methods, the neutron spectrum, irradiation history, and cooling period influence this record. Likewise, the lack of precise nuclear data, such as values of nuclear cross-sections, affects any calculations that can be made. Another difficulty in the non-destructive assay is the presence of high radiation fields which contribute to the ''noise'' background of the measurements. The development of useful and realistic standards is difficult. The non-destructive burn-up methods do serve a useful purpose especially when an approximate value of burn-up is required quickly and economically even though in the present state of the art they lack the desired precision and accuracy. Several non-destructive methods for determining burn-up have been used, are being evaluated, or have been proposed. Various types of spectrometers including the bent crystal, magnetic Compton, Compton coincidence, and scintillation have been used to analyse the gamma radiation from the radioactive material formed during the fission process. Other non-destructive methods include foil activation, neutron transmission, activation analysis, measurement of capture gamma rays, and the measurement of prompt and delayed neutrons. The basic principles of each of the above instruments and methods, their sensitivities and their limitations will be reviewed. Non-destructive methods using stable isotopes produced during the fission process are proposed. In the use of stable isotopes, detailed irradiation history

  10. Experimental results on the MOX fuel. Study of the calculus/measures divergences

    International Nuclear Information System (INIS)

    Martin, S.

    1997-01-01

    For each nuclear plant unit restart, all safety criterion have to be respected. Various parameters as boron concentration, temperature coefficient, worth or power and activity distributions related to fuel assemblies, have to be calculated. To compute these parameters Framatome uses the neutronic channel Science. Before the validation they are compare to experimental measures. For UO 2 fuel the divergence calculus/measures are correct. But for MOX fuels the divergence worsening. This paper discusses tis divergence and research the origin. (A.L.B.)

  11. Chemical characterisation of experimental PFBR-MOX fuel - an overview

    International Nuclear Information System (INIS)

    Mallik, G.K.; Arun Kumar; Panakkal, J.P.; Kamath, H.S.

    2003-01-01

    Fuel pins for experimental PFBR-type fuel containing ( Nat U 0.335 233 U 0.37 Pu 0.295 )O 2 are being manufactured at AFFF and will be irradiated in FBTR. A number of chemical characterization experiments have been carried out on simulated and actual fuel pellets to meet the design specifications of the fuel. An overview is being presented to highlight the related facts. (author)

  12. Design impacts of safeguards and security requirements for a US MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Erkkila, B.H.; Rinard, P.M.; Thomas, K.E.; Zack, N.R.; Jaeger, C.D.

    1998-01-01

    The disposition of plutonium that is no longer required for the nation's defense is being structured to mitigate risks associated with the material's availability. In the 1997 Record of Decision, the US Government endorsed a dual-track approach that could employ domestic commercial reactors to effect the disposition of a portion of the plutonium in the form of mixed oxide (MOX) reactor fuels. To support this decision, the Office of Materials Disposition requested preparation of a document that would review US requirements for safeguards and security and describe their impact on the design of a MOX fuel fabrication facility. The intended users are potential bidders for the construction and operation of the facility. The document emphasizes the relevant DOE Orders but also considers the Nuclear Regulatory Commission (NRC) requirements. Where they are significantly different, the authors have highlighted this difference and provided guidance on the impact to the facility design. Finally, the impacts of International Atomic Energy Agency (IAEA) safeguards on facility design are discussed. Security and materials control and accountability issues that influence facility design are emphasized in each area of discussion. This paper will discuss the prepared report and the issues associated with facility design for implementing practical, modern safeguards and security systems into a new MOX fuel fabrication facility

  13. Fresh MOX fuel transport in Germany: experience for using the MX6

    Energy Technology Data Exchange (ETDEWEB)

    Lallemant, T. [COGEMA Logistics (AREVA Group), Bagnols/sur Ceze (France); Marien, L. [FBFC-I (AREVA Group), Dessel (Belgium); Wagner, R. [RWE, Gundremmingen (Germany); Jahreiss, W. [FRAMATOME ANP GmbH (AREVA Group), Erlangen (Germany); Tschiesche, H. [NCS, Hanau (Germany)

    2004-07-01

    The MX6 packaging developed by COGEMA LOGISTICS replaces the BWR SIEMENS packaging and SIEMENS III packaging for the transport of either BWR or PWR fresh MOX assemblies. It is licensed in France, Germany and Belgium according to TS-R-1 requirements (IAEA 1996). The associated security transport system was developed in co-operation with NCS (Nuclear Cargo + Service GmbH). The MX6 packaging is based on innovative solutions implemented at each step of the design. In 2004, RWE GUNDREMMINGEN Nuclear Power Plant (NPP) will be the first NPP delivered with the MX6 system and MOX assemblies manufactured by BELGONUCLEAIRE and FBFC in Belgium. Before this first transport, successful cold tests were performed for qualification of the whole system with the participation of all parties involved: NPP, carrier, fuel supplier and local Authorities. These tests were conducted by the NPP's operators in FBFC and GUNDREMMINGEN facilities and lead to the validation of the operating manual. Specific conditions for the return of the empty MX6 were also agreed between all parties. Similar operation will be conducted in each NPP before the first use of the MX 6. The large payload of the MX6: - 16 BWR MOX assemblies in one packaging instead of 2 - 6 PWR MOX assemblies in one packaging instead of 3 contributes to the optimisation of the dose uptake during unloading in the NPP. In this paper, the main contributors to the first MOX transport to Germany with the MX6 will present their involvement and feedback at each step of the transport of this new type of packaging, including loading and unloading operations. The use of the MX6 will be extended to other German NPP's from the next year. After FBFC in Belgium, MELOX in France will load the MX6 as well as the current MX8 packaging for the delivery to the French NPP's.

  14. A Computer Simulation to Assess the Nuclear Material Accountancy System of a MOX Fuel Fabrication Facility

    International Nuclear Information System (INIS)

    Portaix, C.G.; Binner, R.; John, H.

    2015-01-01

    SimMOX is a computer programme that simulates container histories as they pass through a MOX facility. It performs two parallel calculations: · the first quantifies the actual movements of material that might be expected to occur, given certain assumptions about, for instance, the accumulation of material and waste, and of their subsequent treatment; · the second quantifies the same movements on the basis of the operator's perception of the quantities involved; that is, they are based on assumptions about quantities contained in the containers. Separate skeletal Excel computer programmes are provided, which can be configured to generate further accountancy results based on these two parallel calculations. SimMOX is flexible in that it makes few assumptions about the order and operational performance of individual activities that might take place at each stage of the process. It is able to do this because its focus is on material flows, and not on the performance of individual processes. Similarly there are no pre-conceptions about the different types of containers that might be involved. At the macroscopic level, the simulation takes steady operation as its base case, i.e., the same quantity of material is deemed to enter and leave the simulated area, over any given period. Transient situations can then be superimposed onto this base scene, by simulating them as operational incidents. A general facility has been incorporated into SimMOX to enable the user to create an ''act of a play'' based on a number of operational incidents that have been built into the programme. By doing this a simulation can be constructed that predicts the way the facility would respond to any number of transient activities. This computer programme can help assess the nuclear material accountancy system of a MOX fuel fabrication facility; for instance the implications of applying NRTA (near real time accountancy). (author)

  15. Fresh MOX fuel transport in Germany: experience for using the MX6

    International Nuclear Information System (INIS)

    Lallemant, T.; Marien, L.; Wagner, R.; Jahreiss, W.; Tschiesche, H.

    2004-01-01

    The MX6 packaging developed by COGEMA LOGISTICS replaces the BWR SIEMENS packaging and SIEMENS III packaging for the transport of either BWR or PWR fresh MOX assemblies. It is licensed in France, Germany and Belgium according to TS-R-1 requirements (IAEA 1996). The associated security transport system was developed in co-operation with NCS (Nuclear Cargo + Service GmbH). The MX6 packaging is based on innovative solutions implemented at each step of the design. In 2004, RWE GUNDREMMINGEN Nuclear Power Plant (NPP) will be the first NPP delivered with the MX6 system and MOX assemblies manufactured by BELGONUCLEAIRE and FBFC in Belgium. Before this first transport, successful cold tests were performed for qualification of the whole system with the participation of all parties involved: NPP, carrier, fuel supplier and local Authorities. These tests were conducted by the NPP's operators in FBFC and GUNDREMMINGEN facilities and lead to the validation of the operating manual. Specific conditions for the return of the empty MX6 were also agreed between all parties. Similar operation will be conducted in each NPP before the first use of the MX 6. The large payload of the MX6: - 16 BWR MOX assemblies in one packaging instead of 2 - 6 PWR MOX assemblies in one packaging instead of 3 contributes to the optimisation of the dose uptake during unloading in the NPP. In this paper, the main contributors to the first MOX transport to Germany with the MX6 will present their involvement and feedback at each step of the transport of this new type of packaging, including loading and unloading operations. The use of the MX6 will be extended to other German NPP's from the next year. After FBFC in Belgium, MELOX in France will load the MX6 as well as the current MX8 packaging for the delivery to the French NPP's

  16. Determination of thorium and plutonium in AHWR experimental (Th, 1%Pu)O2 MOX fuel after microwave dissolution

    International Nuclear Information System (INIS)

    Fulzele, Ajit K.; Malav, R.K.; Pandey, Ashish; Kapoor, Y.S.; Kumar, Manish; Singh, Mamta; Das, D.K.; Prakash, Amrit; Behere, P.G.; Afzal, Mohd

    2013-01-01

    This paper describes determination of thorium and plutonium in experimental (Th, 1%Pu)O 2 AHWR (Advanced Heavy Water Reactor) MOX fuel samples after dissolution by microwave. Time taken to dissolve ∼ 2g of MOX sample by conventional IR heating technique in conc. HNO 3 + 0.05 M HF mixture is about 35-40 hours while using microwave dissolution technique it is ∼ 2 hours. Hence, with the help of microwave dissolution technique analysis time for each sample has been reduced from week to a day. The PuO 2 content (wt%) in the MOX pellets was within specification limit, (1.0±0.1)%. (author)

  17. MOX fuel effective behaviour modeling by a micro-mechanical nonuniform transformation field analysis

    International Nuclear Information System (INIS)

    Largenton, R.

    2012-01-01

    The objective of this research thesis is to develop a modelling by scale change, based on the NTFA approach (Non uniform Transformation Field Analysis). These developments have been achieved on three-dimensional structures which are representative of the MOX fuel, and for local visco-elastic ageing behaviour with free deformations. First, the MOX fuel is represented by using existing methods to process and segment 2D experimental images. 2D information has been upgraded in 3D by a stereo-logic Saltykov method. Tools have been developed to represent and discretize (periodic 3D grid generator) a particulate multiphase composite representative of MOX. Developments made on the NTFA model and on the three-phase particulate composite have been theoretically and numerically studied. The model has then been validated by comparison with reference calculations performed in full field for the effective behaviour as well as for local fields for different test types (imposed strain rate, creep, relaxation, rotating). The approach is then compared with a recently developed homogenisation method: the semi-analytical 'incremental Mori-Tanka' model. Theoretical similarities are outlined. These methods are very fast in terms of CPU time, but the NTFA method remains the one giving the most information, and the most precise, but requires a more important preliminary work (mode identification) [fr

  18. Nuclear materials accountancy in an industrial MOX fuel fabrication plant safeguards versus commercial aspects

    International Nuclear Information System (INIS)

    Canck, H. de; Ingels, R.; Lefevre, R.

    1991-01-01

    In a modern MOX Fuel Fabrication Plant, with a large throughput of nuclear materials, computerized real-time accountancy systems are applied. Following regulations and prescriptions imposed by the Inspectorates EURATOM-IAEA, the State and also by internal plant safety rules, the accountancy is kept in plutonium element, uranium element and 235 U for enriched uranium. In practice, Safeguards Authorities are concerned with quantities of the element (U tot , Pu tot ) and to some extent with its fissile content. Custom Authorities are for historical reasons, interested in fissile quantities (U fiss , Pu fiss ) whereas owners wish to recover the energetic value of their material (Pu equivalent). Balancing the accountancy simultaneously in all these related but not proportional units is a new problem in a MOX-plant where pool accountancy is applied. This paper indicates possible ways to solve the balancing problem created by these different units used for expressing nuclear material quantities

  19. Experience of determination of plutonium and uranium contents in MOX fuel by IDMS

    International Nuclear Information System (INIS)

    Yoshida, Mika; Suzuki, Toru; Kobayashi, Hideo; Ohtani, Tetsuo

    2001-01-01

    In the Plutonium Fuel Center (PFC) of JNC, Isotope Dilution Mass Spectrometry (IDMS) has been used to determine Pu and U contents of nuclear materials since 1996. In MOX fabrication plant, many types of sample with wide variation of Pu/U ratio including aged Pu and process scrap should be analyzed for not only quality control purpose but also material accountancy. Because IDMS can eliminate influences of coexistence elements and has high accuracy, it is considered to be the best analytical method for MOX fabrication plant. This paper summarizes the experience of IDMS in the PFC laboratory including the preparation of Large Size Dried (LSD) spike, and also describes the evaluation of analytical error and consideration on procurement of LSD spike for IDMS

  20. Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

    International Nuclear Information System (INIS)

    Wachs, G.W.

    1997-11-01

    The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)

  1. Chemical analyses and calculation of isotopic compositions of high-burnup UO{sub 2} fuels and MOX fuels

    Energy Technology Data Exchange (ETDEWEB)

    Matsumura, Tetsuo; Sasahara, Akihiro [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

    2001-08-01

    Chemical analysis activities of isotopic compositions of high-burnup UO{sub 2} fuels and MOX fuels in CRIEPI and calculation evaluation are reviewed briefly. C/E values of ORIGEN2, in which original libraries and JENDL-3.2 libraries are used, and other codes with chemical analysis data are reviewed and evaluated. Isotopic compositions of main U and Pu in fuels can be evaluated within 10% relative errors by suitable libraries and codes. Void ratio is effective parameter for C/E values in BWR fuels. JENDL-3.2 library shows remarkable improvement compared with original libraries in isotopic composition evaluations of FP nuclides. (author)

  2. Lessons learned from MELOX plant operation and support to design of new MOX fuel fabrication plants

    International Nuclear Information System (INIS)

    Tourre, Joel; Gattegno, Robert; Guay, Philippe; Bariteau, Jean-Pierre

    2005-01-01

    AREVA is participating in the design of the US MOX Fuel Fabrication Facility (MFFF). To support this project and allow the U.S. Department of Energy (DOE) client to reap full benefit from the MELOX operating experience, AREVA, through COGEMA and its engineering subsidiary SGN have implemented a rigorous process to prudently apply MELOX Lessons Learned to the MFFF design. This paper describes the Lessons Learned process, how the process supports the advancement of fuel fabrication technology and, how the results of the process are benefiting the client. (author)

  3. A review of the thermophysical properties of MOX and UO2 fuels

    International Nuclear Information System (INIS)

    Carbajo, Juan J.; Yoder, Gradyon L.; Popov, Sergey G.; Ivanov, Victor K.

    2001-01-01

    A critical review of the thermophysical properties of UO 2 and MOX fuels has been completed, and the best correlations for thermophysical properties have been selected. The properties reviewed are solidus and liquidus temperatures of the uranium/plutonium dioxide system (melting and solidification temperatures), thermal expansion and density, enthalpy and specific heat, enthalpy (or heat) of fusion, and thermal conductivity. Only fuel properties have been reviewed. The selected set of property correlations was compiled to be used in thermal-hydraulic codes to perform safety calculations

  4. Modernization of RTC for fabrication of MOX fuel, Vibropac fuel pins and BN-600 FA with weapon grade plutonium

    International Nuclear Information System (INIS)

    Grachyov, A.F.; Kalygin, V.V.; Skiba, O.V.; Mayorshin, A. A.; Bychkov, A.V.; Kisly, V.A.; Ovsyannikov, Y.F.; Bobrov, D.A.; Mamontov, S.I.; Tsyganov, A.N.; Churutkin, E.I.; Davydov, P.I.; Samosenko, E.A; Shalak, A.R.; Ojima, Hisao

    2004-01-01

    Since mid 70's RIAR has been performing activities on plutonium involvement in fuel cycle. These activities are considered a stage within the framework of the closed fuel cycle development. Developed at RIAR fuel cycle is based on two technologies: 'dry' process of fuel reprocessing and vibro-packing method for fuel pin fabrication. Due to the available scientific capabilities and a gained experience in operating the technological facilities (ORYOL, SIC) for plutonium (various grade) blending into fuel for fast reactors, RIAR is a participant of the activities aimed at solving these tasks. Under international program RIAR with financial support of JNC (Japan) is modernizing the facility for granulated fuel production, vibro-pac fuel pins and FA fabrication to provide the BN-600 'hybrid' core. In order to provide 'hybrid' core it is necessary to produce (per year): - 1775 kg of granulated MOX-fuel, 6500 fuel pins, 50 fuel assemblies. Potential output of the facility under construction is as follows: - 1800 kg of granulated MOX-fuel per year, 40 fuel pins per shift, 200 FAs for the BN-600 reactor per year. Taking into account domestic and foreign experience in MOX-fuel production, different options were discussed of the equipment layouts in the available premises of chemical technological division of RIAR: - in the shielded manipulator boxes, in the existing hot cells. During construction of the facility in the building under operation the following requirements should be met: - facility must meet all standards and regulations set for nuclear facilities, installation work at the facility must not influence other production programs implemented in the building, engineering supply lines of the facility must be connected to the existing service lines of the building, cost of the activities must not exceed amount of JNC funding. The paper presents results of comparison between two options of the process equipment layout: in boxes and hot cells. This equipment is intended

  5. SRS MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    O'Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site(SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. SRS has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 2 or 3 facility with storage of bulk PuO 2 and assembly, storage, and shipping of fuel bundles in an S and S Category 1 facility. The total Category 1 approach, which is the recommended option, would be done in the 221-H Canyon Building. A facility that was never in service will be removed from one area, and a hardened wall will be constructed in another area to accommodate execution of the LA fuel fabrication. The non-Category 1 approach would require removal of process equipment in the FB-Line metal production and packaging glove boxes, which requires work in a contamination area. The Immobilization Hot Demonstration Program

  6. SRS MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site(SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. SRS has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 2 or 3 facility with storage of bulk PuO{sub 2} and assembly, storage, and shipping of fuel bundles in an S and S Category 1 facility. The total Category 1 approach, which is the recommended option, would be done in the 221-H Canyon Building. A facility that was never in service will be removed from one area, and a hardened wall will be constructed in another area to accommodate execution of the LA fuel fabrication. The non-Category 1 approach would require removal of process equipment in the FB-Line metal production and packaging glove boxes, which requires work in a contamination area. The Immobilization Hot Demonstration Program

  7. Interest in 100% MOX future reactors as seen from the fuel fabrication and from the Pu manager point of view

    International Nuclear Information System (INIS)

    Golinelli, C.; Guillet, J.L.; Nigon, J.L.

    1996-01-01

    Today, plutonium recycling in PWR type reactors has reached the industrial phase. But, on a competitive market, cost reduction can be achieved by improving fuel performances and fuel management. That is why researches on MOX future reactors are still carried out in the world and particularly in France. As a matter of fact, MOX future reactors can be more competitive if the in-reactor utilization is improved. This solution should certainly be the next step to re-use the recovered plutonium from reprocessed spent fuel. (O.M.)

  8. Development of database system on MOX fuel for water reactors (I)

    International Nuclear Information System (INIS)

    Kikuchi, Keiichi; Nakazawa, Hiroaki; Abe, Tomoyuki; Shirai, Takao

    2000-04-01

    JNC has been conducted a great number of irradiation tests to develop MOX fuels for Advanced Thermal Reactor and Light Water Reactors. In order to manage irradiation data consistently and to effectively utilize valuable data obtained from the irradiation tests, we commenced construction of database system on MOX fuel for water reactors in 1998 JFY. Collection and selection of irradiation data and relevant fuel fabrication data, design of the database system and preparation of assisting programs have been finished and data registration onto the system is under way according to priority at present. The database system can be operated through the menu screen on PC. About 94,000 records of data on 11 fuel assemblies in total have been registered onto the database up to the present. By conducting registration of the remaining data and some modification of the system, if necessary, the database system is expected to complete in 2000 JFY. The completed database system is to be distributed to relevant sections in JNC by means of CD-R as a media. This report is an interim report covering 1998 and 1999 JFY, which gives the structure explanation and users manual concerning to the prepared database up to the present. (author)

  9. Transport of fresh MOX fuel assemblies for the Monju initial core

    International Nuclear Information System (INIS)

    Kurakami, J.; Ouchi, Y.; Usami, M.

    1997-01-01

    Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this package design feature such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying of the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule. (Author)

  10. Recent advances in the chemical quality control of MOX fuel for PFBR

    International Nuclear Information System (INIS)

    Prakash, Amrit; Das, D.K.; Behere, P.G.; Afzal, Mohd

    2012-01-01

    Uranium-plutonium mixed oxide (MOX) fuel for Prototype Fast Breeder Reactor (PFBR) is being fabricated at Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre (BARC),Tarapur. A number of quality control steps are required to ensure the quality of the fuel. Chemical characterization of the fuel is very important from reactor performance point of view. More than three hundred batches have been analysed till to date for various specifications like percentage composition, heavy metal content, oxygen to metal ratio, trace metallic impurities, trace non-metallic impurities, cover gas content, total gas content, homogeneity test etc. During these analyses by recommended techniques, studies were carried out to see the feasibility of using methodologies which can reduce the total analysis time, convenience/safety in operation and man rem problems. The present paper describes a glimpse of those studies carried out

  11. International Atomic Energy Agency (IAEA) Activity on Technical Influence of High Burnup UOX and MOX Water Reactor Fuel on Spent Fuel Management

    International Nuclear Information System (INIS)

    Lovasic, Z.; Einziger, R.

    2009-01-01

    This paper briefly reviews the results of the International Atomic Energy Agency (IAEA) project investigating the influence of high burnup and mixed-oxide (MOX) fuels, from water power reactors, on spent fuel management. These data will provide information on the impacts, regarding spent fuel management, for those countries operating light-water reactors (LWR)s and heavy-water reactors (HWR)s with zirconium alloy-clad uranium dioxide (UOX) fuels, that are considering the use of higher burnup UOX or the introduction of reprocessing and MOX fuels. The mechanical designs of lower burnup UOX and higher burnup UOX or MOX fuel are very similar, but some of the properties (e.g., higher fuel rod internal pressures; higher decay heat; higher specific activity; and degraded cladding mechanical properties of higher burnup UOX and MOX spent fuels) may potentially significantly affect the behavior of the fuel after irradiation. These properties are reviewed. The effects of these property changes on wet and dry storage, transportation, reprocessing, re-fabrication of fuel, and final disposal were evaluated, based on regulatory, safety, and operational considerations. Political and strategic considerations were not taken into account since relative importance of technical, economic and strategic considerations vary from country to country. There will also be an impact of these fuels on issues like non-proliferation, safeguards, and sustainability, but because of the complexity of factors affecting those issues, they are only briefly discussed. Data gaps were also identified during this investigation. The pros and cons of using high burnup UOX or MOX, for each applicable issue in each stage of the back end of the fuel cycle, were evaluated and are discussed.. Although, in theory, higher burnup fuel and MOX fuels mean a smaller quantity of spent fuel, the potential need for some changes in design of spent fuel storage, transportation, handling, reprocessing, re-fabrication, and

  12. System analysis of nuclear safety of VVER reactor with MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Klimov, A.D.; Zharkov, V.P.; Suslov, I.R. [Russia, Moscow Malaya Krasnoselskaya St. (Russian Federation); Boyarinov, V.F.; Kevrolev, V.V.; Tchibinyaev, A.V.; Tsibulskiy, V.F. [RRC KI, Russia, Moscow (Russian Federation); Kochurov, B.P. [ITEP, Russia, Moscow (Russian Federation); Giovanni, B. [NFPSC, FRAMATOME (France)

    2005-07-01

    The report presents a short summary of the results achieved in the ISTC (International Science and Technology Center) project 'System analysis of nuclear safety of VVER reactor with MOX fuel' (April 2005). The studies within the project are of a systematic character and include the solutions of 15 tasks. The report gives an overview of the major blocks of these tasks: neutron transport equation solution; calculations of isotopic vectors, analysis of the impact of uncertainties on predicted reactor functionals. The calculation methods, the verification results and the corresponding codes are briefly described. (authors)

  13. Prediction analysis of dose equivalent responses of neutron dosemeters used at a MOX fuel facility

    International Nuclear Information System (INIS)

    Tsujimura, N.; Yoshida, T.; Takada, C.

    2011-01-01

    To predict how accurately neutron dosemeters can measure the neutron dose equivalent (rate) in MOX fuel fabrication facility work environments, the dose equivalent responses of neutron dosemeters were calculated by the spectral folding method. The dosemeters selected included two types of personal dosemeter, namely a thermoluminescent albedo neutron dosemeter and an electronic neutron dosemeter, three moderator-based neutron survey meters, and one special instrument called an H p (10) monitor. The calculations revealed the energy dependences of the responses expected within the entire range of neutron spectral variations observed in neutron fields at workplaces. (authors)

  14. Role of ion chromatography in the chemical characterization of PFBR MOX fuel

    International Nuclear Information System (INIS)

    Kelkar, Anoop; Das, D.K.; Prakash, Amrit; Behere, P.G.; Afzal, Mohd

    2012-01-01

    Ion chromatography (IC) is multi-element technique with the feasibility of determination of metallic as well as non metallic impurities on a single instrument. IC has been used for various analytical purposes in nuclear industry. lt has advantages of low capital investment, small sample size, less radioactive waste generation, comparable precision to spectroscopic techniques and ease of fume hood/glove box adaptation. Present paper describes the determination of trace metallic (alkali, alkaline earth, transition and lanthanide metal ions) and non metallic impurities in PFBR MOX fuel

  15. A plan of reactor physics experiments for reduced-moderation water reactors with MOX fuel in TCA

    International Nuclear Information System (INIS)

    Shimada, Shoichiro; Akie, Hiroshi; Suzaki, Takenori; Okubo, Tutomu; Usui, Shuji; Shirakawa, Toshihisa; Iwamura, Takamiti; Kugo, Teruhiko; Ishikawa, Nobuyuki

    2000-06-01

    The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors which aim at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. Critical Experiments performed so far in Eualope and Japan were reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of the equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX fuel rods used in the experiments are obtained by calculations and the modification of the equipment for the experiments are shown. (author)

  16. International safeguards for a modern MOX [mixed-oxide] fuel fabrication facility

    International Nuclear Information System (INIS)

    Pillay, K.K.S.; Stirpe, D.; Picard, R.R.

    1987-03-01

    Bulk-handling facilities that process plutonium for commercial fuel cycles offer considerable challenges to nuclear materials safeguards. Modern fuel fabrication facilities that handle mixed oxides of plutonium and uranium (MOX) often have large inventories of special nuclear materials in their process lines and in storage areas for feed and product materials. In addition, the remote automated processing prevalent at new MOX facilities, which is necessary to minimize radiation exposures to personnel, tends to limit access for measurements and inspections. The facility design considered in this study incorporates all these features as well as state-of-the-art measurement technologies for materials accounting. Key elements of International Atomic Energy Agency (IAEA) safeguards for such a fuel-cycle facility have been identified in this report, and several issues of primary importance to materials accountancy and IAEA verifications have been examined. We have calculated detection sensitivities for abrupt and protracted diversions of plutonium assuming a single materials balance area for all processing areas. To help achieve optimal use of limited IAEA inspection resources, we have calculated sampling plans for attributes/variables verification. In addition, we have demonstrated the usefulness of calculating σ/sub (MUF-D)/ and detection probabilities corresponding to specified material-loss scenarios and resource allocations. The data developed and the analyses performed during this study can assist both the facility operator and the IAEA in formulating necessary safeguards approaches and verification procedures to implement international safeguards for special nuclear materials

  17. International safeguards for a modern MOX (mixed-oxide) fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Pillay, K.K.S.; Stirpe, D.; Picard, R.R.

    1987-03-01

    Bulk-handling facilities that process plutonium for commercial fuel cycles offer considerable challenges to nuclear materials safeguards. Modern fuel fabrication facilities that handle mixed oxides of plutonium and uranium (MOX) often have large inventories of special nuclear materials in their process lines and in storage areas for feed and product materials. In addition, the remote automated processing prevalent at new MOX facilities, which is necessary to minimize radiation exposures to personnel, tends to limit access for measurements and inspections. The facility design considered in this study incorporates all these features as well as state-of-the-art measurement technologies for materials accounting. Key elements of International Atomic Energy Agency (IAEA) safeguards for such a fuel-cycle facility have been identified in this report, and several issues of primary importance to materials accountancy and IAEA verifications have been examined. We have calculated detection sensitivities for abrupt and protracted diversions of plutonium assuming a single materials balance area for all processing areas. To help achieve optimal use of limited IAEA inspection resources, we have calculated sampling plans for attributes/variables verification. In addition, we have demonstrated the usefulness of calculating sigma/sub (MUF-D)/ and detection probabilities corresponding to specified material-loss scenarios and resource allocations. The data developed and the analyses performed during this study can assist both the facility operator and the IAEA in formulating necessary safeguards approaches and verification procedures to implement international safeguards for special nuclear materials.

  18. Analysis of boiling water reactors capacities for the 100% MOX fuel recycling

    International Nuclear Information System (INIS)

    Knoche, Dietrich

    1999-01-01

    The electro-nuclear park exploitation leads to plutonium production. The plutonium recycling in boiling water reactors performs a use possibility. The difference between the neutronic characteristics of the uranium and the plutonium need to evaluate the substitution impact of UOX fuel by MOX fuel on the reactor operating and safety. The analysis of the main points reached to the following conclusions: the reactivity coefficients are negative, during a cooling accident the re-divergence depends on the isotopic vector of the used plutonium, the efficiency lost of control cross resulting from the plutonium utilization can be compensate by the increase of the B 4C enrichment by 10 B and the change of the steel structure by an hafnium structure, the reactivity control in evolution can be obtained by the fuel poisoning (gadolinium, erbium) and the power map control by the plutonium content monitoring. (A.L.B.)

  19. Non destructive nuclear measurements for control and characterization purpose

    International Nuclear Information System (INIS)

    Lyoussi, Abdallah

    2002-01-01

    In this report for accreditation to supervise researches, the author proposes a large and rather precise overview of his research works which dealt with the upstream and downstream parts of the nuclear fuel cycle. After having discussed the different needs associated with non destructive nuclear measurements during the fuel cycle, the author describes his past research activities. In the following parts, he discusses control and characterization methods associated with the upstream and downstream parts of the fuel cycle: fuel density variation measurement, non destructive control of uranium-235 content of enriched uranium ingots, examination of induced photo-fissions in radioactive waste parcels, use of electron accelerator for simultaneous neutron and photon examination, measurement of the spatial distribution of the photonic component from the Mini Linatron, association of non destructive measurement techniques

  20. Responses of commercially available neutron electronic personal dosemeters in neutron fields simulating workplaces at MOX fuel fabrication facilities

    International Nuclear Information System (INIS)

    Tsujimura, N.; Yoshida, T.; Takada, C.

    2011-01-01

    The authors investigated the performance of three commercially available electronic personal dosemeters (EPDs) in evaluating neutron dose equivalents and discussed their suitability to work environments in MOX fuel fabrication facilities. The EPDs selected for this study were NRY21 (Fuji Electric Systems), PDM-313 (Aloka) and DMC 2000 GN (MGP Instruments). All tests were conducted in moderated 252 Cf neutron fields with neutron spectral and dosimetric characteristics similar to those found in MOX fuel facilities. The test results revealed trends and the magnitude of response variations in relation to neutron spectral changes expected in work environments.

  1. Fuel component of electricity generation cost for the BN-800 reactor with 800 MOX fuel and uranium oxide fuel, increased fuel burnup, and removal of radial breeding blanket

    International Nuclear Information System (INIS)

    Raskach, A.

    2000-01-01

    There are two completed design concepts of NPP with BN-800 type reactors developed with due regard for enhanced safety requirements. They have been created for the 3 rd unit of Beloyarsk NPP and for three units of South Ural NPP. Both concepts are proposed to use mixed oxide fuel (MOX) based on civil plutonium. At this moment economical estimations carried out for these projects need to be revised in connection with the changes of economical situation in Russia and the world nuclear market structure. It is also essential to take into account the existing problem of the excess ex-weapons plutonium utilization and the possibility of using this plutonium to fabricate MOX fuel for the BN-800 reactors. (authors)

  2. Strategy for decommissioning of the glove-boxes in the Belgonucleaire Dessel MOX fuel fabrication plant

    International Nuclear Information System (INIS)

    Vandergheynst, Alain; Cuchet, Jean-Marie

    2007-01-01

    Available in abstract form only. Full text of publication follows: BELGONUCLEAIRE has been operating the Dessel plant from the mid-80's at industrial scale. In this period, over 35 metric tons of plutonium (HM) was processed into almost 100 reloads of MOX fuel for commercial West-European Light Water Reactors. In late 2005, the decision was made to stop the production because of the shortage of MOX fuel market remaining accessible to BELGONUCLEAIRE after the successive capacity increases of the MELOX plant (France) and the commissioning of the SMP plant (UK). As a significant part of the decommissioning project of this Dessel plant, about 170 medium-sized glove-boxes are planned for dismantling. In this paper, after having reviewed the different specifications of ±-contaminated waste in Belgium, the authors introduce the different options considered for cleaning, size reduction and packaging of the glove-boxes, and the main decision criteria (process, α-containment, mechanization and radiation protection, safety aspects, generation of secondary waste, etc) are analyzed. The selected strategy consists in using cold cutting techniques and manual operation in shielded disposable glove-tents, and packaging α-waste in 200-liter drums for off-site conditioning and intermediate disposal. (authors)

  3. Nuclear terrorism risk analysis using game theory. Case study of sea transportation of MOX fuel

    International Nuclear Information System (INIS)

    Nakatani, Eri; Tanaka, Satoru; Choi, Jor-Shan

    2010-01-01

    While considerable attention and resources have been directed towards improving nuclear security in Japan in response to the threat of nuclear terrorism, the transport of nuclear material raises concern by the public as indicated in the recent return of MOX fuel from Europe. This concern cannot be adequately addressed by the government through communications with the public because of the confidential nature of such transport. Also, it remains a challenge for adequately assessing the nuclear terrorism risk because many key parameters associated with such assessment cannot be derived from statistical data and reflect actors' intentions unlike assessment on natural disasters. This study proposes an assessment methodology which introduces game theory to deduce the correlations between those key parameters and can be used to analyze the nuclear terrorism risk, both quantitatively and qualitatively for the civilian use of nuclear power. Risk will be calculated by Monte Carlo methods based on probability distributions set for actors' utilities. A case-study of transporting the MOX fuel by sea is also included. (author)

  4. LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

    Energy Technology Data Exchange (ETDEWEB)

    Bronson, M.C.

    1997-10-01

    The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium.

  5. Sensitivity and uncertainty analysis for UO2 and MOX fueled PWR cells

    International Nuclear Information System (INIS)

    Foad, Basma; Takeda, Toshikazu

    2015-01-01

    Highlights: • A method for calculating sensitivity coefficients has been improved. • The IR approximation was used in order to get accurate results. • Sensitivities and uncertainties are calculated using the improved method. • The method is applied for UO 2 and MOX fueled PWR cells. • The verification was performed by comparing our results with MCNP6 and TSUNAMI-1D. - Abstract: This paper discusses the improvement of a method for calculating sensitivity coefficients of neutronics parameters relative to infinite dilution cross-sections because the conventional method neglects resonance self-shielding effect. In this study, the self-shielding effect is taken into account by using the intermediate resonance approximation in order to get accurate results in both high and low energy groups. The improved method is applied to calculate sensitivity coefficients and uncertainties of eigenvalue responses for UO 2 and MOX (ThO 2 –UO 2 and PuO 2 –UO 2 ) fueled pressurized water reactor cells. The verification of the improved method was performed by comparing the sensitivities with MCNP6 and TSUNAMI-1D. For uncertainty, calculation comparisons were done with TSUNAMI-1D, and we demonstrate that the differences are caused by the use of different covariance matrices

  6. LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

    International Nuclear Information System (INIS)

    Bronson, M.C.

    1997-01-01

    The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium

  7. Modelling the actual behaviour of the MOX fuel by a micromechanical analysis in non-uniform transformation fields

    International Nuclear Information System (INIS)

    Largenton, R.

    2012-01-01

    This research thesis aimed at developing a model based on scale change to assess more precisely the distribution of local thermo-mechanical fields within a heterogeneous medium as MOX fuel. The analysis method is a non-uniform transformation field analysis (NTFA) which is adapted to the problem of scale change in presence of a coupling between dissipative and elastic effects. More precisely, the author addressed the development of a NTFA model based on specific three-phase and three-dimensional microstructures which are typical of the MOX fuel in an in-service operation. The first part proposes an overview of knowledge and use of MOX. It recalls the context and the industrial problematic associated with this fuel: operating principles for a 900 MWe PWR, fuel fabrication processes, fuel morphologies and structural and microstructural consequences. It addresses local mechanisms within each phase during irradiation, and presents the approach methodology regarding scale change. The second part reports the representation and analysis in complete fields of multiphase particle-based composites (MOX type) in order to determine the representative elementary volume and the local behaviour of each phase. The third part reports the extension of the NTFA approach to 3D aspects, free deformations, ageing and optimization. The last part compares the NTFA approach with the incremental two-phase and three-phase Mori-Tanaka models

  8. Verification of the depletion capabilities of the MCNPX code on a LWR MOX fuel assembly

    International Nuclear Information System (INIS)

    Cerba, S.; Hrncir, M.; Necas, V.

    2012-01-01

    The study deals with the verification of the depletion capabilities of the MCNPX code, which is a linked Monte-Carlo depletion code. For such a purpose the IV-B phase of the OECD NEA Burnup credit benchmark has been chosen. The mentioned benchmark is a code to code comparison of the multiplication coefficient k eff and the isotopic composition of a LWR MOX fuel assembly at three given burnup levels and after five years of cooling. The benchmark consists of 6 cases, 2 different Pu vectors and 3 geometry models, however in this study only the fuel assembly calculations with two Pu vectors were performed. The aim of this study was to compare the obtained result with data from the participants of the OECD NEA Burnup Credit project and confirm the burnup capability of the MCNPX code. (Authors)

  9. Analysis of Radial Plutonium Isotope Distribution in Irradiated Test MOX Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jae Yong; Lee, Byung Ho; Koo, Yang Hyun; Kim, Han Soo

    2009-01-15

    After Rod 3 and 6 (KAERI MOX) were irradiated in the Halden reactor, their post-irradiation examinations are being carried out now. In this report, PLUTON code was implemented to analyze Rod 3 and 6 (KAERI MOX). In the both rods, the ratio of a maximum burnup to an average burnup in the radial distribution was 1.3 and the contents of {sup 239}Pu tended to increase as the radial position approached the periphery of the fuel pellet. The detailed radial distribution of {sup 239}Pu and {sup 240}Pu, however, were somewhat different. To find the reason for this difference, the contents of Pu isotopes were investigated as the burnup increased. The content of {sup 239}Pu decreased with the burnup. The content of {sup 240}Pu increased with the burnup by the 20 GWd/tM but decreased over the 20 GWd/tM. The local burnup of Rod 3 is higher than that of Rod 6 due to the hole penetrated through the fuel rod. The content of {sup 239}Pu decreased more rapidly than that of {sup 240}Pu in the Rod 6 with the increased burnup. It resulted in a radial distribution of {sup 239}Pu and {sup 240}Pu similar to Rod 3. The ratio of Xe to Kr is a parameter to find where the fissions occur in the nuclear fuel. In both Rod 3 and 6, it was 18.3 in the whole fuel rod cross section, which showed that the fissions occurred in the plutonium.

  10. The MOX fuel behaviour test IFA-597.4/.5/.6/.7; Summary of in-pile fuel temperature and gas release data

    Energy Technology Data Exchange (ETDEWEB)

    Koike, Hisashi

    2003-11-15

    It is considered important to study the in-reactor behaviour of MOX fuel in order to enhance the database on such fuel. For this reason, IFA-597.4/.5/.6/.7 were included in the joint research programme of the Halden Project. The series of tests, containing two MIMAS-MOX fuel rods, both equipped with a fuel centre thermocouple and a pressure bellows transducer, has been irradiated in the Halden Reactor since July 1997 under HBWR conditions. The objectives of the test series were to study the thermal and fission gas release (FGR) behaviour of MOX fuel and to explore potential differences in behaviour between solid and hollow pellets. One of the rods had mainly solid pellets, while the other contained only hollow pellets. Both rods had an initial Pu-fissile enrichment of 6.07%. The cladding outside diameter was 9.50 mm, and the initial fuel-clad gap was 180 mum. In the course of the test, power upratings for FGR studies of the MOX fuel were planned at burnup intervals of about 10 MWd/kg MOX. The power uprating was successfully performed at approx10 MWd/kg MOX, where the estimated fuel peak temperature of the solid pellets exceeded the FGR threshold temperature for UO{sub 2} fuel, while that of the hollow pellets remained below the threshold. For the solid fuel, the temperature at onset of FGR was consistent with the empirical threshold temperature for UO{sub 2} fuel. For the hollow fuel, gas release was observed at temperatures below the threshold. FGRs at the end-of-life were approx17% for the solid pellet rod and approx14% for the hollow pellet rod, respectively. As a result of discussions in HPG meetings, IFA-597.7 was unloaded in January 2002. PIE was carried out to check in-pile pressure measurements and examine fuel structural characteristics. The discharge burn-up of the MOX fuel was 32 MWd/kg MOX as determined from in-pile power data. This report supersedes HWR-712 (June 2002) previously issued on in-pile data from IFA-597.4/5/6/7. (Author)

  11. ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    O'Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1997-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO 2 and UO 2 ), typically containing 95% or more UO 2 . DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement

  12. ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1997-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement.

  13. M4/12 package project - development of a package for transport of new MOX fuel in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Kaye, B.R.; Porter, I.; Ashley, P. [BNFL, Warrington, Cheshire (United Kingdom)

    2004-07-01

    BNFL has a requirement to deliver new MOX fuel from the Sellafield MOX Plant (SMP) to its customers in mainland Europe. To satisfy this requirement, a transport system has been developed which complies with national and international regulations and conventions relating to the transport of Category 1 materials. Fundamental to this system is the transport package. BNFL has designed, developed, and is manufacturing a new transport package, the M4/12, This paper gives a brief overview of the overall transport system and then goes on to describe the development of the M4/12 package with particular emphasis on the novel features of the design.

  14. M4/12 package project - development of a package for transport of new MOX fuel in Europe

    International Nuclear Information System (INIS)

    Kaye, B.R.; Porter, I.; Ashley, P.

    2004-01-01

    BNFL has a requirement to deliver new MOX fuel from the Sellafield MOX Plant (SMP) to its customers in mainland Europe. To satisfy this requirement, a transport system has been developed which complies with national and international regulations and conventions relating to the transport of Category 1 materials. Fundamental to this system is the transport package. BNFL has designed, developed, and is manufacturing a new transport package, the M4/12, This paper gives a brief overview of the overall transport system and then goes on to describe the development of the M4/12 package with particular emphasis on the novel features of the design

  15. Development of hotcell non-destructive examination techniques

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Uhn; Yu, S. C.; Kang, B. S.; Byun, K. S. [Chungbuk National University, Chungju (Korea)

    2002-01-01

    The purpose of this project is to establish non-destructive examination techniques which needs to determine the status of spent nuclear fuel and/or bundles. Through the project, we will establish an image reconstruction tomography which is a kind of non-destructive techniques in Hotcell. The tomography technique can be used to identify the 2-dimensional density distribution of fission products in the spent fuel rods and/or bundles. And form results of the measurement and analysis of magnetic properties of neutron irradiated material in the press vessel and reactor, we will develop some techniques to test its hardness and defects. In 2001, the first year, we have established mathematical background and necessary data and informations to develop the techniques. We will try to find some experimental results that are necessary in developing the Hotcell non-destructive examination techniques in the coming year. 14 refs., 65 figs., 5 tabs. (Author)

  16. OPT-TWO: Calculation code for two-dimensional MOX fuel models in the optimum concentration distribution

    International Nuclear Information System (INIS)

    Sato, Shohei; Okuno, Hiroshi; Sakai, Tomohiro

    2007-08-01

    OPT-TWO is a calculation code which calculates the optimum concentration distribution, i.e., the most conservative concentration distribution in the aspect of nuclear criticality safety, of MOX (mixed uranium and plutonium oxide) fuels in the two-dimensional system. To achieve the optimum concentration distribution, we apply the principle of flattened fuel importance distribution with which the fuel system has the highest reactivity. Based on this principle, OPT-TWO takes the following 3 calculation steps iteratively to achieve the optimum concentration distribution with flattened fuel importance: (1) the forward and adjoint neutron fluxes, and the neutron multiplication factor, with TWOTRAN code which is a two-dimensional neutron transport code based on the SN method, (2) the fuel importance, and (3) the quantity of the transferring fuel. In OPT-TWO, the components of MOX fuel are MOX powder, uranium dioxide powder and additive. This report describes the content of the calculation, the computational method, and the installation method of the OPT-TWO, and also describes the application method of the criticality calculation of OPT-TWO. (author)

  17. Fabrication and characterization of Am, Np and Cm bearing MOX fuel obtained by conventional powder metallurgy

    Energy Technology Data Exchange (ETDEWEB)

    Jankowiak, A.; Leorier, C.; Desmouliere, F.; Donnet, L. [Commissariat a l' Energie Atomique (CEA), CEA/DEN/VRH/DTEC/SDTC/LEMA, 30207 Bagnols-sur-Ceze cedex (France)

    2008-07-01

    Transmutation of minor actinides enables to produce energy and to turn them into shorter-lived nuclides. This promising way to reduce the long-term waste radiotoxicity is world wide investigated. In the framework of the Global Actinide Cycle International Demonstration and regarding the homogeneous recycling for transmutation in fast reactors, minor actinides (Am, Np, Cm) bearing MOX fuel pellets were fabricated in the ATALANTE facility by a conventional powder metallurgy process (milling then pressing and finally sintering). The sintered pellets were submitted to a visual inspection where neither crack nor strain was detected. In addition, the pellets exhibit a density in the range 93-96% TD which makes them proper to the irradiation in fast reactors. The pellets were characterized by XRD (X radiation diffraction) and SEM (scanning electron microscopy) combined to image analysis. (authors)

  18. Improved MOX fuel calculations using new Pu-239, Am-241 and Pu-240 evaluations

    International Nuclear Information System (INIS)

    Noguere, G.; Bouland, O.; Bernard, D.; Leconte, P.; Blaise, P.; Peneliau, Y.; Vidal, J.F.; Saint Jean, C. de; Leal, L.; Schilleebeeckx, P.; Kopecky, S.; Lampoudis, C.

    2013-01-01

    Several studies based on the JEFF-3.1.1 nuclear data library show a systematic over-estimation of the critical keff for core configurations of MOX fuel assemblies. The present work investigates possible improvements of the C/E results by using new evaluations for Am-241, Pu-239 and Pu-240. The work reported in this paper demonstrates the performances of the new Am-241 evaluation based on capture and transmission data measured at the IRMM. For Pu-239, the new evaluation, established in the frame of the WPEC/SG-34, is able to explain a systematic discrepancy observed between different EOLE experiments. The combination of the Am-241 and Pu-239 evaluations demonstrates the necessity to improve the radiation width of the first resonance of Pu-240

  19. Experiences with the first prototype MOX fuel rods fabricated at Argentina

    International Nuclear Information System (INIS)

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

    1996-01-01

    The irradiation of the first Argentine prototypes of pressurized heavy water reactor (PHWR) (U,Pu)O sub 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 alpha Facility. 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

  20. Sensitivity and uncertainty analysis of reactivities for UO2 and MOX fueled PWR cells

    Energy Technology Data Exchange (ETDEWEB)

    Foad, Basma [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan); Egypt Nuclear and Radiological Regulatory Authority, 3 Ahmad El Zomar St., Nasr City, Cairo, 11787 (Egypt); Takeda, Toshikazu [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan)

    2015-12-31

    The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO{sub 2} and MOX fueled pressurized water reactor cells. The improved method has been used to calculate sensitivity coefficients relative to infinite dilution cross-sections, where the self-shielding effect is taken into account. Two types of reactivities are considered: Doppler reactivity and coolant void reactivity, for each type of reactivity, the sensitivities are calculated for small and large perturbations. The results have demonstrated that the reactivity responses have larger relative uncertainty than eigenvalue responses. In addition, the uncertainty of coolant void reactivity is much greater than Doppler reactivity especially for large perturbations. The sensitivity coefficients and uncertainties of both reactivities were verified by comparing with SCALE code results using ENDF/B-VII library and good agreements have been found.

  1. Specific application of burnup credit for MOX PWR fuels in the rotary dissolver

    International Nuclear Information System (INIS)

    Caplin, Gregory; Coulaud, Alexandre; Klenov, Pavel; Toubon, Herve

    2003-01-01

    In prospect of a Mixed OXide spent fuels processing in the rotary dissolver in COGEMA/La Hague plant, it is interesting to quantify the criticality-safety margins from the burnup credit. Using the current production computer codes and considering a minimal fuel irradiation of 3 200 megawatt-day per ton, this paper shows the impact of burnup credit on industrial parameters such as the permissible concentration in the dissolution solution or the permissible oxide mass in the rotary dissolver. Moreover, the burnup credit is broken down into five sequences in order to quantify the contribution of fissile nuclides decrease and of minor actinides and fission products formation. The implementation of the burnup credit in the criticality-safety analysis of the rotary dissolver may lead to workable industrial conditions for the particular MOX fuel studied. It can eventually be noticed that minor actinides contribution is negligible and that considering only the six major fission products is sufficient, owing to the weak fuel irradiation contemplated. (author)

  2. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  3. Simulation of the neutron-physical properties of the classical UO2 fuel and of MOX fuel during the burn-up by Transuranus

    International Nuclear Information System (INIS)

    Breza, J. jr.; Necas, V.; Daoeilek, P.

    2005-01-01

    The classical nuclear fuel UO 2 is well known for VVER reactors. Nevertheless, in the near future it will be possible to replace this fuel by novel, advanced kinds of fuel, for instance MOX, inert matrices fuel, etc., that will allow to increase the level of burn-up and minimize the amount of hazardous waste. The code Transuranus [2], designed at ITU Karlsruhe, is intended for thermal and mechanical analyses of fuel elements in nuclear reactors. We have utilized the code Transuranus to simulate the neutron-physical properties of the classical UO 2 fuel and of MOX fuel during the burn-up to a level of 40 MWd/kgHM. We compare obtained results of uranium and plutonium nuclides concentrations, their changes during burn-up, with results obtained by code HELIOS [3], which is well-validated code for this kind of applications. We performed calculations of fission gasses concentrations, namely xenon and krypton. (author)

  4. Full MOX core design in ABWR

    International Nuclear Information System (INIS)

    Ihara, Toshiteru; Mochida, Takaaki; Izutsu, Sadayuki; Fujimaki, Shingo

    2003-01-01

    Electric Power Development Co., Ltd. (EPDC) has been investigating an ABWR plant for construction at Oma-machi in Aomori Prefecture. The reactor, termed FULL MOX-ABWR will have its reactor core eventually loaded entirely with mixed-oxide (MOX) fuel. Extended use of MOX fuel in the plant is expected to play important roles in the country's nuclear fuel recycling policy. MOX fuel bundles will initially be loaded only to less than one-third of the reactor, but will be increased to cover its entire core eventually. The number of MOX fuel bundles in the core thus varies anywhere from 0 to 264 for the initial cycle and, 0 to 872 for equilibrium cycles. The safety design of the FULL MOX-ABWR briefly stated next considers any probable MOX loading combinations out of such MOX bundle usage scheme, starting from full UO 2 to full MOX cores. (author)

  5. Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

    International Nuclear Information System (INIS)

    El bakkari, B.; El Bardouni, T.; Merroun, O.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Chakir, E.

    2009-01-01

    The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc...). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called 'BUCAL1'. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k ∞ ) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.

  6. Development of an integrated, unattended assay system for LWR-MOX fuel pellet trays

    International Nuclear Information System (INIS)

    Stewart, J.E.; Hatcher, C.R.; Pollat, L.L.

    1994-01-01

    Four identical unattended plutonium assay systems have been developed for use at the new light-water-reactor mixed oxide (LWR-MOX) fuel fabrication facility at Hanau, Germany. The systems provide quantitative plutonium verification for all MOX pellet trays entering or leaving a large, intermediate store. Pellet-tray transport and storage systems are highly automated. Data from the ''I-Point'' (information point) assay systems will be shared by the Euratom and International Atomic Energy Agency (IAEA) Inspectorates. The I-Point system integrates, for the first time, passive neutron coincidence counting (NCC) with electro-mechanical sensing (EMS) in unattended mode. Also, provisions have been made for adding high-resolution gamma spectroscopy. The system accumulates data for every tray entering or leaving the store between inspector visits. During an inspection, data are analyzed and compared with operator declarations for the previous inspection period, nominally one month. Specification of the I-point system resulted from a collaboration between the IAEA, Euratom, Siemens, and Los Alamos. Hardware was developed by Siemens and Los Alamos through a bilateral agreement between the German Federal Ministry of Research and Technology (BMFT) and the US DOE. Siemens also provided the EMS subsystem, including software. Through the USSupport Program to the IAEA, Los Alamos developed the NCC software (NCC COLLECT) and also the software for merging and reviewing the EMS and NCC data (MERGE/REVIEW). This paper describes the overall I-Point system, but emphasizes the NCC subsystem, along with the NCC COLLECT and MERGE/REVIEW codes. We also summarize comprehensive testing results that define the quality of assay performance

  7. Study of the lattice parameter evolution of PWR irradiated MOX fuel by X-Ray diffraction

    International Nuclear Information System (INIS)

    Clavier, B.

    1995-01-01

    Fuel irradiation leads to a swelling resulting from the formation of gaseous (Kr, Xe) or solid fission products which are found either in solution or as solid inclusions in the matrix. This phenomena has to be evaluated to be taken into account in fuel cladding Interaction. Fuel swelling was studied as a function of burn up by measuring the corresponding cell constant evolution by X-Ray diffraction. This study was realized on Mixed Oxide Fuels (MOX) irradiated in a Pressurized Water Reactor (PWR) at different burn-up for 3 initial Pu contents. Lattice parameter evolutions were followed as a function of burn-up for the irradiated fuel with and without an annealing thermal treatment. These experimental evolutions are compared to the theoretical evolutions calculated from the hard sphere model, using the fission product concentrations determined by the APPOLO computer code. Contribution of varying parameters influencing the unit cell value is discussed. Thermal treatment effects were checked by metallography, X-Ray diffraction and microprobe analysis. After thermal treatment, no structural change was observed but a decrease of the lattice parameter was measured. This modification results essentially from self-irradiation defect annealing and not from stoichiometry variations. Microprobe analysis showed that about 15% of the formed Molybdenum is in solid solution In the oxide matrix. Micrographs showed the existence of Pu packs in the oxide matrix which induces a broadening of diffraction lines. The RIETVELD method used to analyze the X-Ray patterns did not allow to characterize independently the Pu packs and the oxide matrix lattice parameters. Nevertheless, with this method, the presence of micro-strains in the irradiated nuclear fuel could be confirmed. (author)

  8. Mimas, a mature and flexible process to convert the stockpiles of separated civil and weapon grade plutonium into MOX fuel for use in LWR's

    International Nuclear Information System (INIS)

    Vandergheynst, A.; Vanderborck, Y.

    2001-01-01

    The BELGONUCLEAIRE Dessel MOX fabrication plant started operation in 1973. The first ten years have laid down the bases for all the modifications and improvements in the field of fuel fabrication and quality control process and technology, waste management, safety and safeguards. In 1984, BELGONUCLEAIRE developed the MIMAS fabrication process and has used it on industrial scale to make MOX fuel complying with the most stringent fuel vendor specifications. From 1986 to 2000, more than 25 t Pu have been processed into more than 450 tHM of MIMAS fuel delivered in five countries. The MOX fuel produced has been demonstrated to reach at least the same performance as the UO 2 fuel used simultaneously in the same reactors. The BELGONUCLEAIRE MIMAS MOX fuel fabrication process was selected by COGEMA in the late 80(tm)s for its MELOX and its Cadarache plants. In 1999, the MIMAS process was chosen by the US DOE for the new MOX fabrication plant to be built in Savannah (SC-USA) to ''demilitarize'' 25,6 tons of weapon grade plutonium originating from nuclear war- heads. Recently MIMAS was selected by Japan for its domestic MOX plant to be built in Rokkasho-mura. (author)

  9. The state of the art in non destructive testing of nuclear fuel cladding tubes using ultrasonic rotary systems; on line computer and statistics

    International Nuclear Information System (INIS)

    Rauscher, Rudolf

    Nondestructive evaluation of nuclear fuel cladding by ultrasonic tests is described. Ultrasonic transducers for detection of flaws and dimensions are built in a rotary system with a speed of 8000 rpm. The testing system is adapted to a configuration consisting of two microcomputers connected to each other

  10. Non-destructive testing at Chalk River

    International Nuclear Information System (INIS)

    Hilborn, J.W.

    1976-01-01

    In 1969 CRNL recognized the need for a strong group skilled in non-destructive test procedures. Within two years a new branch called Quality Control Branch was staffed and working. This branch engages in all aspects of non-destructive testing including development of new techniques, new applications of known technology, and special problems in support of operating reactors. (author)

  11. Oxidizing dissolution of spent MOX47 fuel subjected to water radiolysis: Solution chemistry and surface characterization by Raman spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Jegou, C., E-mail: christophe.jegou@cea.f [Commissariat a l' Energie Atomique (CEA), Marcoule Reasearch Center, B.P. 17171, F-30207 Bagnols-sur-Ceze Cedex (France); Caraballo, R.; De Bonfils, J.; Broudic, V.; Peuget, S. [Commissariat a l' Energie Atomique (CEA), Marcoule Reasearch Center, B.P. 17171, F-30207 Bagnols-sur-Ceze Cedex (France); Vercouter, T. [Commissariat a l' Energie Atomique (CEA), Saclay Reasearch Center, B.P. 11, F-91191 Gif-sur-Yvette Cedex (France); Roudil, D. [Commissariat a l' Energie Atomique (CEA), Marcoule Reasearch Center, B.P. 17171, F-30207 Bagnols-sur-Ceze Cedex (France)

    2010-04-01

    The mechanisms of oxidizing dissolution of spent MOX fuel (MIMAS TU2 (registered) ) subjected to water radiolysis were investigated experimentally by leaching spent MOX47 fuel samples in pure water at 25 deg. C under different oxidizing conditions (with and without external gamma irradiation); the leached surfaces were characterized by Raman spectroscopy. The highly oxidizing conditions resulting from external gamma irradiation significantly increased the concentration of plutonium (Pu(V)) and uranium (U(VI)) compared with a benchmark experiment (without external irradiation). The oxidation behavior of the plutonium-enriched aggregates differed significantly from that of the UO{sub 2} matrix after several months of leaching in water under gamma irradiation. The plutonium in the aggregates appears to limit fuel oxidation. The only secondary phases formed and identified to date by Raman spectroscopy are uranium peroxides that generally precipitate on the surface of the UO{sub 2} grains. Concerning the behavior of plutonium, solution analysis results appear to be compatible with a conventional explanation based on an equilibrium with a Pu(OH){sub 4(am)} phase. The fission product release - considered as a general indicator of matrix alteration - from MOX47 fuel also increases under external gamma irradiation and a change in the leaching mode is observed. Diffusive leaching was clearly identified, coinciding with the rapid onset of steady-state actinide concentrations in the bulk solution.

  12. A model for the description of the evolution of PU agglomerates in MOX fuels

    Energy Technology Data Exchange (ETDEWEB)

    Federici, E [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France); Blanpain, P [FRAMATOME, Lyon (France); Permezel, P [Electricite de France, Moret-sur-Loing (France)

    1997-08-01

    In order to describe the irradiation behavior of Pu agglomerates under LWR steady state conditions in MIMAS MOX fuels, a model including the neutronic evolution of the heavy atoms and their diffusion processes between the agglomerates and the matrix has been developed. It leads to the calculations of Pu enrichment in the two phases and of the agglomerates size evolution during irradiation. The calculated distribution of the fission in the fuel gives access to the local power and burnup heterogeneity factor. Electron probe microanalyses (EPMA) have been carried out on fuels irradiated up to 45000 MWd/tM. Diametral and local distribution of Pu are used to calculate the enrichments of the agglomerates and the matrix, which are then compared to the results of the model. During irradiation, the Pu concentration falls markedly in the agglomerates and increases steadily in the matrix, leading to a homogenization of the fuel on a microstructural scale. Heterogeneity factors give an estimate of the deviation from homogeneity. Knowing the local fission rate and burnup in the agglomerates and the matrix enables the calculation of the local fission gas concentrations, which are compared to the xenon EPMA diametral distribution. Comparison with the calculated matrix xenon concentration at the edge of the pellet where there is no gas release, shows that some fission gas atoms which originated from the agglomerates, have been dissolved in the matrix by recoil. The calculated gas concentrations give an estimate of the quantity of gas dissolved. This work has been performed with the intent to improved fuel rod performance code estimates of fission gas concentrations retained or released in both the matrix and the agglomerates. (author). 4 refs, 7 figs.

  13. A model for the description of the evolution of PU agglomerates in MOX fuels

    International Nuclear Information System (INIS)

    Federici, E.; Blanpain, P.; Permezel, P.

    1997-01-01

    In order to describe the irradiation behavior of Pu agglomerates under LWR steady state conditions in MIMAS MOX fuels, a model including the neutronic evolution of the heavy atoms and their diffusion processes between the agglomerates and the matrix has been developed. It leads to the calculations of Pu enrichment in the two phases and of the agglomerates size evolution during irradiation. The calculated distribution of the fission in the fuel gives access to the local power and burnup heterogeneity factor. Electron probe microanalyses (EPMA) have been carried out on fuels irradiated up to 45000 MWd/tM. Diametral and local distribution of Pu are used to calculate the enrichments of the agglomerates and the matrix, which are then compared to the results of the model. During irradiation, the Pu concentration falls markedly in the agglomerates and increases steadily in the matrix, leading to a homogenization of the fuel on a microstructural scale. Heterogeneity factors give an estimate of the deviation from homogeneity. Knowing the local fission rate and burnup in the agglomerates and the matrix enables the calculation of the local fission gas concentrations, which are compared to the xenon EPMA diametral distribution. Comparison with the calculated matrix xenon concentration at the edge of the pellet where there is no gas release, shows that some fission gas atoms which originated from the agglomerates, have been dissolved in the matrix by recoil. The calculated gas concentrations give an estimate of the quantity of gas dissolved. This work has been performed with the intent to improved fuel rod performance code estimates of fission gas concentrations retained or released in both the matrix and the agglomerates. (author). 4 refs, 7 figs

  14. Options for converting excess plutonium to feed for the MOX fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Watts, Joe A [Los Alamos National Laboratory; Smith, Paul H [Los Alamos National Laboratory; Psaras, John D [Los Alamos National Laboratory; Jarvinen, Gordon D [Los Alamos National Laboratory; Costa, David A [Los Alamos National Laboratory; Joyce, Jr., Edward L [Los Alamos National Laboratory

    2009-01-01

    The storage and safekeeping of excess plutonium in the United States represents a multibillion-dollar lifecycle cost to the taxpayers and poses challenges to National Security and Nuclear Non-Proliferation. Los Alamos National Laboratory is considering options for converting some portion of the 13 metric tons of excess plutonium that was previously destined for long-term waste disposition into feed for the MOX Fuel Fabrication Facility (MFFF). This approach could reduce storage costs and security ri sks, and produce fuel for nuclear energy at the same time. Over the course of 30 years of weapons related plutonium production, Los Alamos has developed a number of flow sheets aimed at separation and purification of plutonium. Flow sheets for converting metal to oxide and for removing chloride and fluoride from plutonium residues have been developed and withstood the test oftime. This presentation will address some potential options for utilizing processes and infrastructure developed by Defense Programs to transform a large variety of highly impure plutonium into feedstock for the MFFF.

  15. Report on Evaluation, Verification, and Assessment of Porosity Migration Model in Fast Reactor MOX Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Novascone, Stephen Rhead [Idaho National Lab. (INL), Idaho Falls, ID (United States); Peterson, John William [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Abstract This report documents the progress of simulating pore migration in ceramic (UO2 and mixed oxide or MOX) fuel using BISON. The porosity field is treated as a function of space and time whose evolution is governed by a custom convection-diffusion-reaction equation (described here) which is coupled to the heat transfer equation via the temperature field. The porosity is initialized to a constant value at every point in the domain, and as the temperature (and its gradient) are increased by application of a heat source, the pores move up the thermal gradient and accumulate at the center of the fuel in a time-frame that is consistent with observations from experiments. There is an inverse dependence of the fuel’s thermal conductivity on porosity (increasing porosity decreases thermal conductivity, and vice-versa) which is also accounted for, allowing the porosity equation to couple back into the heat transfer equation. Results from an example simulation are shown to demonstrate the new capability.

  16. Validation of the TUBRNP model with the radial distribution of plutonium in MOX fuel measured by SIMS and EPMA

    Energy Technology Data Exchange (ETDEWEB)

    O` Carroll, C; Laar, J Van De; Walker, C T [CEC Joint Research Centre, Karlsruhe (Germany)

    1997-08-01

    The new model TUBRNP (TRANSURANUS burnup) predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of plutonium. Comparisons between measurements and the prediction of the TUBRNP model have been made for UO{sub 2} LWR fuels: they were found to be in excellent agreement and it is seen that TUBRNP is a marked improved on previous models. A powerful techniques for the characterization of irradiation fuel is Electron Probe Microanalysis (EPMA). Uranium, plutonium and fission product distributions can be analysed quantitatively. A complement, providing isotopic information with a lateral resolution comparable to EPMA, is secondary ion mass spectrometry (SIMS). Recently, the technique has been successfully applied for the measurement of the radial distribution of plutonium isotopes in irradiated nuclear fuel pins. The extension of the TUBRNP model to mixed oxide fuels seems to be the natural step to take. In MOX fuels the picture is greatly complicated by the presence of the (U, Pu)O{sub 2} agglomerates. The rim effect referred to above may be masked by the high concentrations of plutonium in the bulk of the fuel. A detailed investigation of a number of MOX fuel samples has been made using the TUBRNP model. Results are presented for a range of fuels with different enrichment and burnup. Through its participation in the PRIMO and DOMO programmes, PSI in conjunction with the Institute for Transuranium Elements had the opportunity to validate the new theoretical model TUBRNP. The authors with therefore to express their thanks to the organizers and to the numerous European and Japanese organizations which have supported these two international programmes on MOX fuel behavior. 7 refs, 9 figs, 3 tabs.

  17. LWR high burn-up operation and MOX introduction. Fuel cycle performance from the viewpoint of waste management

    International Nuclear Information System (INIS)

    Inagaki, Yaohiro; Iwasaki, Tomohiko; Niibori, Yuichi; Sato, Seichi; Ohe, Toshiaki; Kato, Kazuyuki; Torikai, Seishi; Nagasaki, Shinya; Kitayama, Kazumi

    2009-01-01

    From the viewpoint of waste management, a quantitative evaluation of LWR nuclear fuel cycle system performance was carried out, considering both higher burn-up operation of UO 2 fuel coupled with the introduction of MOX fuel. A major parameter to quantify this performance is the number of high-level waste (HLW) glass units generated per GWd (gigawatt-day based on reactor thermal power generation before electrical conversion). This parameter was evaluated for each system up to a maximum burn-up of 70GWd/THM (gigawatt-day per ton of heavy metal) assuming current conventional reprocessing and vitrification conditions where the waste loading of glass is restricted by the heat generation rate, the MoO 3 content, or the noble metal content. The results showed that higher burn-up operation has no significant influence on the number of glass units generated per GWd for UO 2 fuel, though the number of glass units per THM increases linearly with burn-up and is restricted by the heat generation rate. On the other hand, the introduction of MOX fuel causes the number of glass units per GWd to double owing to the increase in the heat generation rate. An extended cooling period of the spent fuel prior to reprocessing effectively reduces the heat generation rate for UO 2 fuel, while a separation of minor actinides (Np, Am, and Cm) from the high-level waste provides additional reduction for MOX fuel. However, neither of these leads to a substantial reduction in the number of glass units, since the MoO 3 content or the noble metal content restricts the number of glass units rather than the heat generation rate. These results suggest that both the MoO 3 content and the noble metal content provide the key to reducing the amount of waste glass that is generated, leading to an overall improvement in fuel cycle system performance. (author)

  18. A technical review of non-destructive assay research for the characterization of spent nuclear fuel assemblies being conducted under the US DOE NGSI

    Energy Technology Data Exchange (ETDEWEB)

    Croft, Stephen [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

    2010-12-06

    There is a growing belief that expansion of nuclear energy generation will be needed in the coming decades as part of a mixed supply chain to meet global energy demand. At stake is the health of the economic engine that delivers human prosperity. As a consequence renewed interest is being paid to the safe management of spent nuclear fuel (SNF) and the plutonium it contains. In addition to being an economically valuable resource because it can be used to construct explosive devices, Pu must be placed on an inventory and handled securely. A multiinstitutional team of diverse specialists has been assembled under a project funded by the US Department of Energy (DOE) Next Generation Safeguards Initiative (NGSI) to address ways to nondestructively quantify the plutonium content of spent nuclear fuel assemblies, and to also detect the potential diversion of pins from those assemblies. Studies are underway using mostly Monte Carlo tools to assess the feasibility, individual and collective performance capability of some fourteen nondestructive assay methods. Some of the methods are familiar but are being applied in a new way against a challenging target which is being represented with a higher degree of realism in simulation space than has been done before, while other methods are novel. In this work we provide a brief review of the techniques being studied and highlight the main achievements to date. We also draw attention to the deficiencies identified in for example modeling capability and available basic nuclear data. We conclude that this is an exciting time to be working in the NDA field and that much work, both fundamental and applied, remains ahead if we are to advance the state of the practice to meet the challenges posed to domestic and international safeguards by the expansion of nuclear energy together with the emergence of alternative fuel cycles.

  19. A technical review of non-destructive assay research for the characterization of spent nuclear fuel assemblies being conducted under the US DOE NGSI

    International Nuclear Information System (INIS)

    Croft, Stephen; Tobin, Stephen J.

    2010-01-01

    There is a growing belief that expansion of nuclear energy generation will be needed in the coming decades as part of a mixed supply chain to meet global energy demand. At stake is the health of the economic engine that delivers human prosperity. As a consequence renewed interest is being paid to the safe management of spent nuclear fuel (SNF) and the plutonium it contains. In addition to being an economically valuable resource because it can be used to construct explosive devices, Pu must be placed on an inventory and handled securely. A multiinstitutional team of diverse specialists has been assembled under a project funded by the US Department of Energy (DOE) Next Generation Safeguards Initiative (NGSI) to address ways to nondestructively quantify the plutonium content of spent nuclear fuel assemblies, and to also detect the potential diversion of pins from those assemblies. Studies are underway using mostly Monte Carlo tools to assess the feasibility, individual and collective performance capability of some fourteen nondestructive assay methods. Some of the methods are familiar but are being applied in a new way against a challenging target which is being represented with a higher degree of realism in simulation space than has been done before, while other methods are novel. In this work we provide a brief review of the techniques being studied and highlight the main achievements to date. We also draw attention to the deficiencies identified in for example modeling capability and available basic nuclear data. We conclude that this is an exciting time to be working in the NDA field and that much work, both fundamental and applied, remains ahead if we are to advance the state of the practice to meet the challenges posed to domestic and international safeguards by the expansion of nuclear energy together with the emergence of alternative fuel cycles.

  20. An experimental investigation of accumulation and transmutation behavior of americium in the MOX fuel irradiated in a fast reactor

    International Nuclear Information System (INIS)

    Osaka, Masahiko; Koyama, Shin-ichi; Maeda, Shigetaka; Mitsugashira, Toshiaki

    2005-01-01

    Americium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Americium was isolated from the irradiated MOX fuel by a combined method of anion-exchange chromatography and oxidation of Am. The isotopic ratios of americium and its content were determined by thermal ionization mass spectroscopy and α-spectrometry, respectively. The americium isotopic ratio was similar for all the specimens, but was significantly different from that of PWR-MOX. On the basis of present analytical results, the accumulation and transmutation behavior of americium nuclides in a fast reactor is discussed from the viewpoints of neutron spectrum dependence and the isomeric ratio of the 241 Am capture reaction. The estimated isomeric ratio is about 87%, which is close to the latest evaluated value. A rapid estimation method of Am content by using the 240 Pu to 239 Pu ratio was adopted and proved to be valid for the spent fuel irradiated in the fast reactor

  1. Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ito, Masahiro; Maeda, Koji [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan)

    2011-09-30

    Highlights: > We evaluated diametral strain of fast reactor MOX fuel pins irradiated to 130 GWd/t. > The strain was due to cladding void swelling and irradiation creep. > The irradiation creep was caused by internal gas pressure and PCMI. > The PCMI was associated with pellet swelling by rim structure or by cesium uranate. > The latter effect tended to increase the cumulative damage fraction of the cladding. - Abstract: The C3M irradiation test, which was conducted in the experimental fast reactor, 'Joyo', demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, 'Monju'. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and {sup 137}Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

  2. Present status of reactor physics in the United States and Japan-IV. 2. Micro-Reactor Physics of MOX-Fueled Core

    International Nuclear Information System (INIS)

    Takeda, Toshikazu

    2001-01-01

    Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design. We used the subgroup method to treat the space dependence of the self-shielding effect of heavy nuclides, and we used the characteristics method to treat the angular dependence of neutron flux in a fuel pellet. Figure 1 compares the power distributions in MOX and UO 2 fuel cells at the beginning of burnup. The power is calculated with and without considering the space dependence of the self-shielding effect of the cross sections. For the MOX cell, the power distribution has a peak at the cell edge because of large Pu absorption especially when considering the spatial self-shielding effect. When a MOX rod is adjacent to UO 2 fuel rods, the flux distribution has an azimuthal dependence in addition to the radial dependence within a rod. For example, consider a 2x2 fuel assembly composed of three UO 2 rods and one MOX rod, with the mirror reflection boundary condition. A burnup calculation was done with the condition; the radius of the MOX pellet is divided into two regions, and the azimuthal angle is divided into eight. The number density of 239 Pu at 44 000 MWd/t for the MOX rod shows azimuthal dependence by 20%. The maximum burnup occurs in the direction of the UO 2 rods. This is

  3. Investigation of fiberoptic behaviour during gamma irradiation. Part of a coordinated programme on non-destructive techniques for reactor fuel characterization

    International Nuclear Information System (INIS)

    Boeck, H.

    1981-03-01

    Radiation induced effects in glass-rods and fiberoptics have been studied to determine parameters affecting the application of these materials in endoscopes operating in severe radiation environments such as the core of a nuclear power plant or the spent fuel element storage pool. Different glass and fiber types have been investigated and various transmission properties were found. The samples were irradiated in the thermal column of the TRIGA Mark II reactor, Vienna. The radiation induced transmission loss was measured approximately 1 hour after the reactor shutdown. Saturation effects in the transmission loss were observed and changes in glass and fiber recovery vs time as a function of radiation dose and photo bleaching methods were studied. Photo bleaching was investigated with a quartz lamp, an arc lamp and an UV-laser light. (author)

  4. Performance values for non destructive assay (NDA) techniques applied to safeguards: the 2002 evaluation by the ESARDA NDA Working Group

    International Nuclear Information System (INIS)

    Guardini, S.

    2003-01-01

    The first evaluation of NDA performance values undertaken by the ESARDA Working Group for Standards and Non Destructive Assay Techniques (WGNDA) was published in 1993. Almost 10 years later the Working Group decided to review those values, to report about improvements and to issue new performance values for techniques which were not applied in the early nineties, or were at that time only emerging. Non-Destructive Assay techniques have become more and more important in recent years, and they are used to a large extent in nuclear material accountancy and control both by operators and control authorities. As a consequence, the performance evaluation for NDA techniques is of particular relevance to safeguards authorities in optimising Safeguards operations and reducing costs. Performance values are important also for NMAC regulators, to define detection levels, limits for anomalies, goal quantities and to negotiate basic audit rules. This paper presents the latest evaluation of ESARDA Performance Values (EPVs) for the most common NDA techniques currently used for the assay of nuclear materials for Safeguards purposes. The main topics covered by the document are: techniques for plutonium bearing materials: PuO 2 and MOX; techniques for U-bearing materials; techniques for U and Pu in liquid form; techniques for spent fuel assay. This issue of the performance values is the result of specific international round robin exercises, field measurements and ad hoc experiments, evaluated and discussed in the ESARDA NDA Working Group. (author)

  5. Can non-destructive inspection be reliable

    International Nuclear Information System (INIS)

    Silk, M.G.; Stoneham, A.M.; Temple, J.A.G.

    1988-01-01

    The paper on inspection is based on the book ''The reliability of non-destructive inspection: assessing the assessment of structures under stress'' by the present authors (published by Adam Hilger 1987). Emphasis is placed on the reliability of inspection and whether cracks in welds or flaws in components can be detected. The need for non-destructive testing and the historical attitudes to non-destructive testing are outlined, along with the case of failure. Factors influencing reliable inspection are discussed, and defect detection trials involving round robin tests are described. The development of reliable inspection techniques and the costs of reliability and unreliability are also examined. (U.K.)

  6. Estimate of the instant release fraction for UO2 and MOX fuel at t=0

    International Nuclear Information System (INIS)

    Johnson, L.; Poinssot, C; Ferry, C.; Lovera, P.

    2004-07-01

    values, which results in significant overprediction of average IRF values. Best estimate IRF values are determined for moderate burnup UO 2 fuel for nuclides for which data exist, because the understanding and data is sufficient. Only pessimistic IRF values are estimated for radionuclides for which little data is available and in the case of MOX fuel and higher burnup UO 2 fuel. Special attention is given to several phenomena occurring in the outer region of fuel pellets (rim region) resulting in restructuring of fuel grains. These include: a) high fission density as a result of high yields of 239 Pu arising from capture of epithermal neutrons; b) increased porosity; c) reduction in grain size; d) increased thermal release of fission gas from the grains. From the perspective of assessing the release of fission products from spent fuel under disposal conditions, the restructuring process is important

  7. 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

  8. OECD benchmark a of MOX fueled PWR unit cells using SAS2H, triton and mocup

    International Nuclear Information System (INIS)

    Ganda, F.; Greenspan, A.

    2005-01-01

    Three code systems are tested by applying them to calculate the OECD PWR MOX unit cell benchmark A. The codes tested are the SAS2H code sequence of the SCALE5 code package using 44 group library, MOCUP (MCNP4C + ORIGEN2), and the new TRITON depletion sequence of SCALE5 using 238 group cross sections generated using CENTRM with continuous energy cross sections. The burnup-dependent k ∞ and actinides concentration calculated by all three code-systems were found to be in good agreement with the OECD benchmark average results. Limited results were calculated also with the WIMS-ANL code package. WIMS-ANL was found to significantly under-predict k ∞ as well as the concentration of Pu 242 , consistently with the predictions of the WIMS-LWR reported by two of the OECD benchmark participants. Additionally, SAS2H is benchmarked against MOCUP for a hydride fuel containing unit cell, giving very satisfactory agreement. (authors)

  9. Validation studies on quick analysis of MOX fuel by combination of laser induced breakdown spectroscopy and ablation resonance absorption spectroscopy

    International Nuclear Information System (INIS)

    Wakaida, Ikuo; Akaoka, Katsuaki; Miyabe, Masabumi; Kato, Masaaki; Otobe, Haruyoshi; Ohoba, Hironori; Khumaeni, Ali

    2014-01-01

    Research and development of laser based quick analysis without chemical analysis and neutron measurement for next-generation Minor Actinide containing MOX fuel has been carried out, and the basic performances by using un-irradiated MOX fuel were demonstrated. The glove box had been re-constructed and specialized for laser spectroscopy, and the remote spectroscopy of MOX sample contained several concentrations of Pu was performed. In elemental analysis by Laser Induced Breakdown Spectroscopy (LIBS) with high resolution spectrometer, relative error of 2.9% at 30% Pu and the detection lower limit of 2500ppm in natural U oxide were demonstrated with the operation time of 5 min. In isotope ratio analysis by Ablation Resonance Absorption Spectroscopy, tunable semiconductor laser system was constructed, and the performances such as relative deviation less than 1% in the ratio of "2"4"0Pu/"2"3"9Pu and the sensitivity of 30-100ppm in natural U were also accomplished with laser operation time of 3 to 5min. As for an elemental analysis of the simulated liquid sample, ultra-thin laminate flow was experimented as LIBS target, and the sensitivity comparable to conventional ICP-AES was confirmed. Present study includes the result of the entrusted project by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT). (author)

  10. Non Destructive Analysis of Uranium by Radiometry

    International Nuclear Information System (INIS)

    Yusuf Nampira

    2007-01-01

    Uranium used in nuclear fuel development activity. the Substance use incurred by regulation safeguard. On that account in uranium acceptance conducted by verification of according to document by the specification of goods. Verification done by analysing performed uranium. The activity require by analyse method which simple and rapid analyses and has accurate result of analyses, is hence done by validation of non destructive uranium analysis that is with count gamma radiation from 235 U and product decay from 238 U. Quantitative analysis of uranium in substance determined by through count radiation-g at energy 185.72 keV and the use assess ratio of gamma radiation count from 235 U to 234 Pa to determine isotope content 235 U in substance. The result of analyses were given result of analysis with above correctness storey level 95% and have limit detect equivalent by 0.0174 mg U in U 3 O 8 . This method use at isotope uranium-235 analysis through count gamma radiation comparing method 235 U/ 234 Pa giving accuracy level 95% at sample equivalent uranium its content in 1 g uranium with isotope 235 U smaller than 75 weight percent. (author)

  11. The new deterministic 3-D radiation transport code Multitrans: C5G7 MOX fuel assembly benchmark

    International Nuclear Information System (INIS)

    Kotiluoto, P.

    2003-01-01

    The novel deterministic three-dimensional radiation transport code MultiTrans is based on combination of the advanced tree multigrid technique and the simplified P3 (SP3) radiation transport approximation. In the tree multigrid technique, an automatic mesh refinement is performed on material surfaces. The tree multigrid is generated directly from stereo-lithography (STL) files exported by computer-aided design (CAD) systems, thus allowing an easy interface for construction and upgrading of the geometry. The deterministic MultiTrans code allows fast solution of complicated three-dimensional transport problems in detail, offering a new tool for nuclear applications in reactor physics. In order to determine the feasibility of a new code, computational benchmarks need to be carried out. In this work, MultiTrans code is tested for a seven-group three-dimensional MOX fuel assembly transport benchmark without spatial homogenization (NEA C5G7 MOX). (author)

  12. Fuel component of electricity generation cost for the BN-800 reactor with MOX fuel and uranium oxide fuel with increasing of fuel burnup and removing of radial breeding blanket

    International Nuclear Information System (INIS)

    Raskach, A.

    2001-01-01

    Nowadays there are two completed design concepts of Nuclear Power Plants (NPPs) with the BN-800 type reactors developed with due regard for advanced safety requirements. One of them is the design of the fourth unit of the Beloyarsk Nuclear Power Plant; the other one is the design of three units of the South Ural Nuclear Power Plant. The both concepts are to use mixed oxide fuel (MOX fuel) based on civil plutonium. Studies on any project include economical analyses and cost of fuel is an essential parameter. In the course of the design works on the both projects such evaluations were done. For BN-800 on the Beloyarsk site nuclear fuel costs were taken from actual expenses of the BN-600 reactor and converted to rated thermal power and design capacity factor of the BN-800 and then increased by 20% in connection with turning to MOX fuel. Then this methodology was rewarding, but the ratio of uranium fuel and MOX fuel costs might change for the last years. For the project of three units of the South Ural Nuclear Power Plant nuclear fuel expenses were calculated from the data on a MOX fuel fabrication production facility (Complex-300). However, investigations performed recently shown that the methodology of economical assessments should be revised, as well as design and technology of MOX fuel fabrication at Complex-300 should be revised to meet all the existing safety requirements. Excepting there is a great bulk of civil plutonium to be reproduced, now we came up against the problem to utilize the exceeding ex-weapons plutonium that obviously can be used for MOX fuel fabrication as well. Construction of the MOX fuel fabrication facility - Complex-300 - was started in 1983. Its design output was planned to provide simultaneously 4 fast reactors of the BN-800 type with MOX fuel. By now about 50% of construction works (taking into account auxiliary buildings and arrangements) and 20% of installation works have been done at Complex-300. Along this, first works to construct

  13. Cr2O3-doped MOX fuel: doping and sintering atmosphere optimization

    International Nuclear Information System (INIS)

    Thomas, R.

    2013-01-01

    Optimal use of the Mixed Oxide (U,Pu)O 2 nuclear fuel in pressurized water reactors is mainly limited by the behavior of gaseous fission produced during irradiation. Within the MOX microstructure, the probability of fission gas release is increased by the presence of rich localized plutonium areas exhibiting a higher local burn-up. A solution consists in optimizing plutonium distribution within the industrial product and promoting the crystalline growth of the fuel grains. For this purpose, addition of chromium sesquioxide during the manufacturing process is currently considered. A previous thesis has shown that the best results are obtained for a Cr addition slightly greater than the solubility limit of Cr in (U,Pu)O 2 . In order to explain the enhanced plutonium homogeneity, the author highlighted the formation of PuCrO 3 precipitates at grain boundaries. A sintering model under reducing atmosphere, with chromium addition, was proposed. However, several points have to be more thoroughly investigated, especially regarding the solubility limit of chromium, as well as the optimal conditions of PuCrO 3 precipitates formation. In a first part, speciation of solubilized and precipitated chromium in the mixed oxide (U,Pu)O 2 is studied using electron probe microanalysis (EPMA) and X-ray absorption spectroscopy (XAS). It was shown that the oxidation state and the environment of soluble chromium within the (U,Pu)O 2 matrix do not depend on the oxygen partial pressure during sintering, neither on the plutonium content of the mixed oxide. However, both chemical nature of the precipitates and chromium solubility depend on the thermodynamic variable and on the plutonium content.Based on these results, a chromium solubility model in the mixed oxide (U,Pu)O 2-x was built using the law of mass action governing solubility equilibrium. This model is described as a function of the plutonium content (y) of the solid solution (U 1-y Pu y )O 2-x (y = 0,11; 0,275 et 1) and in the

  14. Development of MOX manufacturing technology in BNFL

    International Nuclear Information System (INIS)

    Buchan, P.G.; Powell, D.J.; Edwards, J.

    1998-01-01

    BNFL is successfully operating a small scale MOX fuel fabrication facility at its Sellafield Site and is currently constructing an advanced, commercial scale MOX facility to complement its existing LWR UO 2 fabrication capability. BNFL's MOX fuel capability is fully supported by a comprehensive technology development programme aimed at providing a high quality product which is successfully competing in the market. Building on the experience gained over the last 30 years, is from the production of both thermal and fast reactor MOX fuels, BNFL's development team set a standard for its MOX product which is targeted at exceeding the performance of UO 2 fuel in reactor. In order to meet the stringent design requirements the product development team has introduced the Short Binderless Route (SBR) process that is now used routinely in BNFL's MOX Demonstration Facility (MDF) and which forms the basis for BNFL's large scale Sellafield MOX Plant. This plant not only uses the SBR process for MOX production but also incorporates the most advanced technology available anywhere in the world for nuclear fuel production. A detailed account of the technology developed by BNFL to support its MOX fuels business will be provided, together with an explanation of the processes and plants used for MOX fuel production by BNFL. The paper also looks at the future needs of the MOX business and how improvements in pellet design can assist the MOX fabrication production process to meet the user demand requirements of utilities around the world. (author)

  15. Design of full MOX core in ABWR

    International Nuclear Information System (INIS)

    Kinoshita, Y.; Hirose, T.; Sasagawa, M.; Sakuma, T

    1999-01-01

    A Full MOX-ABWR, loaded with mixed-oxide (MOX) fuels of up to 100% of the core, is planned. Increased MOX fuel utilization will result in greater savings of uranium. Studies on the fuel rod thermal-mechanical design, the core design and the safety evaluation have been made, and the results are summarized in this paper. To sum it all up, the safety of the Full MOX-ABWR has been confirmed through design evaluations adequately considering the MOX fuel and core characteristics. (author)

  16. Use of destructive and nondestructive methods of analysis for quality assurance at MOX fuel production in the Russia

    International Nuclear Information System (INIS)

    Bibilashvili, Y.K.; Rudenko, V.S.; Chorokhov, N.A.; Korovin, Y.I.; Petrov, A.M.; Vorobiev, A.V.; Mukhortov, N.F.; Smirnov, Y.A.; Kudryavtsev, V.N.

    2000-01-01

    Parameters of MOX fuel with various plutonium contents are considered from the point of view of necessity of their control for quality assurance. Destructive and nondestructive methods used for this purpose in the Russia are described: controlled potential coulometry for determination of uranium or/and plutonium contents, their ratio and oxygen factor; mass spectrometry for determination of uranium and plutonium isotopic composition; chemical spectral emission method for determination of contents of 'metal' impurities, boron and silicon, and methods of determination of gas forming impurities. Capabilities of nondestructive gamma-ray spectrometry techniques are considered in detail and results of their use at measurement of uranium and plutonium isotopic composition in initial dioxides, at determination of contents of uranium and plutonium, and uniformity of their distribution in MOX powder and pellets. The necessity of correction of algorithm of the MGA program is shown for using the program at analyses of gamma-ray spectra of MOX with low contents of low burnup plutonium. (authors)

  17. Theoretical and Experimental Research in Neutron Spectra and Nuclear Waste Transmutation on Fast Subcritical Assembly with MOX Fuel

    Science.gov (United States)

    Arkhipkin, D. A.; Buttsev, V. S.; Chigrinov, S. E.; Kutuev, R. Kh.; Polanski, A.; Rakhno, I. L.; Sissakian, A.; Zulkarneev, R. Ya.; Zulkarneeva, Yu. R.

    2003-07-01

    The paper deals with theoretical and experimental investigation of transmutation rates for a number of long-lived fission products and minor actinides, as well as with neutron spectra formed in a subcritical assembly driven with the following monodirectional beams: 660-MeV protons and 14-MeV neutrons. In this work, the main objective is the comparison of neutron spectra in the MOX assembly for different external driving sources: a 660-MeV proton accelerator and a 14-MeV neutron generator. The SAD project (JINR, Russia) has being discussed. In the context of this project, a subcritical assembly consisting of a cylindrical lead target surrounded by a cylindrical MOX fuel layer will be constructed. Present conceptual design of the subcritical assembly is based on the core with a nominal unit capacity of 15 kW (thermal). This corresponds to a multiplication coefficient, keff= 0.945, and an accelerator beam power of 0.5 kW. The results of theoretical investigations on the possibility of incinerating long-lived fission products and minor actinides in fast neutron spectrum and formation of neutron spectra with different hardness in subcritical systems based on the MOX subcritical assembly are discussed. Calculated neutron spectra emitted from a lead target irradiated by a 660-MeV protons are also presented.

  18. Glove box adaptation, installation and commissioning of WD-XRF system for determination of PuO2 in MOX fuel samples

    International Nuclear Information System (INIS)

    Aher, Sachin; Pandey, Ashish; Khan, F.A.; Das, D.K.; Kumar, Surendra; Behere, P.G.; Mohd Afzal

    2015-01-01

    Glove Box facility forms the foremost important confinement system at nuclear fuel fabrication facility for handling of Plutonium based MOX fuels. Due to limited resources of Natural Uranium and maximum utilization of thorium, India has adopted 'Close Fuel Cycle Strategy' which involves use of Plutonium based fuels in Thermal and Fast reactors. Plutonium being radio toxic material with a higher biological half-life, Plutonium based fuel fabrication facility requires special techniques and confinement as a primary method for protection against spreading of powder contamination. Glove Box along with dynamic ventilation and HEPA Filters forms the preeminent facility for safe handling of plutonium based MOX fuels. Various equipment's, systems and instruments associated with MOX fuel production are need to be adapted inside the Glove Box with considerations of safety, ergonomics, accessibility for operations and maintenance, connections of various feed through like electrical connections, gas line supply etc. Quality Control plays the vital role in production of MOX fuels to ensure the finest quality of product to meet the defined specifications of MOX fuels. Presently AFFF is fabricating MOX fuel containing 21% and 28% PuO 2 along with DDUO 2 the first core of PFBR. Precise quantification of PuO 2 in MOX fuel pellets is necessary process control steps after batch preparation in Milling and Mixing operation. At AFFF, WD-XRF is one of the system used for determination of percentage of PuO 2 in MOX fuel batch. Glove Box adaptation of WD-XRF system along with 30 Tones Hydraulic press for sample preparation is being carried out in Type VI and Type IV Glove Boxes connected through transfer tunnel. Due to restrictions of space inside the Glove Box, a special mechanism is developed and installed for safe titling of WD-XRF system inside the Glove Box during the need of maintenance. These Glove Boxes are leak tested by various leak testing technique to meet the

  19. Argentine nuclear fuels MOX irradiated in the Petten reactor: Analysis of experience with the BACO code

    Energy Technology Data Exchange (ETDEWEB)

    Marino, A C; Perez, E; Adelfang, P [Argentine Atomic Energy Commission, Buenos Aires (Argentina)

    1997-08-01

    The irradiation of our first prototypes of MOX nuclear fuels fabricated in Argentina began in 1986. These experiences had been made in the HFR-Petten reactor, Holland. The six rods were fabricated in the {alpha} Facility (GAID-CNEA-Argentina). The first rod has been used for destructive pre-irradiation analysis in the KFK (Kernforschungszentrum Karlsruhe), Germany. The second one was a pathfinder for calibrating systems in the HFR. Another two rods included doped pellets based on iodine. One of them included CsI and auxiliary components. The second one included elemental iodine. The concentration of iodine was intended to simulate 15 MWd/ton(M) of burnup. We defined the power histories with the BACO code. We assumed a cycle of 15 days that included interaction treatments of cladding and pellet due to the power cycling. The last ramp is let run until stress corrosion cracking (SCC) is induced. The experience named BU15 was done with the last two rods. The final burnup was 15 MWd/ton(M), and a final ramp test was arranged for one of them. This burnup is the same as the previous two rods. The power level during irradiation was low and without major solicitations, only the normal shutdowns of the HFR. The ramp was similar to that used for the iodine test. We attempt to see the correct correspondence between the BU15 and the doping test. The pathfinder had an excellent behavior in the HFR reactor. The presence of microcracks inside the cladding was observed in the iodine test as we predicted with the BACO code. The post-irradiation tests of the BU15 experience has just ended. The development of the ramp was interrupted due to an increase of activity in the system. We presumed the presence of a failure in the rod. The visual inspection of the rod shows an atypical failure for this kind of fuel, i.e. they found a small circular hole. We use the BACO code for the behavior analysis of the fuel rods. 23 refs, 29 figs, 5 tabs.

  20. Argentine nuclear fuels MOX irradiated in the Petten reactor: Analysis of experience with the BACO code

    International Nuclear Information System (INIS)

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

    1997-01-01

    The irradiation of our first prototypes of MOX nuclear fuels fabricated in Argentina began in 1986. These experiences had been made in the HFR-Petten reactor, Holland. The six rods were fabricated in the α Facility (GAID-CNEA-Argentina). The first rod has been used for destructive pre-irradiation analysis in the KFK (Kernforschungszentrum Karlsruhe), Germany. The second one was a pathfinder for calibrating systems in the HFR. Another two rods included doped pellets based on iodine. One of them included CsI and auxiliary components. The second one included elemental iodine. The concentration of iodine was intended to simulate 15 MWd/ton(M) of burnup. We defined the power histories with the BACO code. We assumed a cycle of 15 days that included interaction treatments of cladding and pellet due to the power cycling. The last ramp is let run until stress corrosion cracking (SCC) is induced. The experience named BU15 was done with the last two rods. The final burnup was 15 MWd/ton(M), and a final ramp test was arranged for one of them. This burnup is the same as the previous two rods. The power level during irradiation was low and without major solicitations, only the normal shutdowns of the HFR. The ramp was similar to that used for the iodine test. We attempt to see the correct correspondence between the BU15 and the doping test. The pathfinder had an excellent behavior in the HFR reactor. The presence of microcracks inside the cladding was observed in the iodine test as we predicted with the BACO code. The post-irradiation tests of the BU15 experience has just ended. The development of the ramp was interrupted due to an increase of activity in the system. We presumed the presence of a failure in the rod. The visual inspection of the rod shows an atypical failure for this kind of fuel, i.e. they found a small circular hole. We use the BACO code for the behavior analysis of the fuel rods. 23 refs, 29 figs, 5 tabs

  1. Evaluation of existing United States' facilities for use as a mixed-oxide (MOX) fuel fabrication facility for plutonium disposition

    International Nuclear Information System (INIS)

    Beard, C.A.; Buksa, J.J.; Chidester, K.; Eaton, S.L.; Motley, F.E.; Siebe, D.A.

    1995-01-01

    A number of existing US facilities were evaluated for use as a mixed-oxide fuel fabrication facility for plutonium disposition. These facilities include the Fuels Material Examination Facility (FMEF) at Hanford, the Washington Power Supply Unit 1 (WNP-1) facility at Hanford, the Barnwell Nuclear Fuel Plant (BNFP) at Barnwell, SC, the Fuel Processing Facility (FPF) at Idaho National Engineering Laboratory (INEL), the Device Assembly Facility (DAF) at the Nevada Test Site (NTS), and the P-reactor at the Savannah River Site (SRS). The study consisted of evaluating each facility in terms of available process space, available building support systems (i.e., HVAC, security systems, existing process equipment, etc.), available regional infrastructure (i.e., emergency response teams, protective force teams, available transportation routes, etc.), and ability to integrate the MOX fabrication process into the facility in an operationally-sound manner that requires a minimum amount of structural modifications

  2. Effect of high burn-up and MOX fuel on reprocessing, vitrification and disposal of PWR and BWR spent fuels based on accurate burn-up calculation

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, T.; Iwasaki, T.; Wada, K. [Tohoku Univ., Graduate School of Engineering, Dept. of Quantum Science and Energy Engineering, Sendai 980-8579 (Japan); Suyama, K. [Japan Atomic Energy Agency, Shirakata-Shirane 2-4, Naka-gun, Ibaraki-ken 319-1195 (Japan)

    2006-07-01

    To examine the procedures of the reprocessing, the vitrification and the geologic disposal, precise burn-up calculation for high burn-up and MOX fuels has been performed for not only PWR but also BWR by using SWAT and SWAT2 codes which are the integrated bum-up calculation code systems combined with the bum-up calculation code, ORIGEN2, and the transport calculation code, SRAC (the collision probability method) or MVP (the continuous energy Monte Carlo method), respectively. The calculation results shows that all of the evaluated items (heat generation and concentrations of Mo and Pt) largely increase and those significantly effect to the current procedures of the vitrification and the geologic disposal. The calculation result by SWAT2 confirms that the bundle calculation is required for BWR to be discussed about those effects in details, especially for the MOX fuel. (authors)

  3. Non-Destructive Testing of Plutonium-Enriched Fuel Rods; Controles non Destructifs de Barreaux Combustibles Enrichis au Plutonium; Nedestruktivnyj-kontrol' toplivnykh sterzhnej''obogashchennykh plutoniem; Ensayo no Destructivo de Elementos Combustibles Enriquecidos en Plutonio

    Energy Technology Data Exchange (ETDEWEB)

    Libotte, P.; Inniger, H.; Leblanc, J. M. [Centre D' Etude de l' Energie Nucleaire, Societe Belge pour l' Industrie Nucleaire, S.A. (Belgonucleaire), Mol (Belgium)

    1965-09-15

    The Societe Belge pour L'Industrie Nucleaire (BelgoNucleaire) and the Centre d'Etude de l'Energie Nucleaire (C.E.N.) are conducting a study under contract and in collaboration with the Commission of EURATOM on the use of plutonium in pressurizedwater power reactors. During these investigations a number of non-destructive testing techniques have been developed for assessing the quality of the fabricated fuel rods. The non-destructive testing techniques described in the paper are used primarily for checking the dimensions and quality of fuel-element cladding tubes, the integrity of welds, the leak-tightness of the rods, the density and distribution of the fuel, and plutonium enrichment and radial and axial distribution in the fuel rod. The principal non-destructive testing methods used are radiography, autoradiography, gamma-radiography and ultrasonics. The authors discuss the principle underlying each procedure and describe the equipment used, together with the techniques applied in each case to obtain the required sensitivity and accuracy. With the radiography technique, flaws of the order of 50-60 {mu}m in the end-cap welds of fuel elements can be detected. Since the fuel consists of three high-density powder fractions compacted by vibration, with only the fine fraction comprising the plutonium enrichment, it is essential to measure accurately the distribution of the plutonium and the density distribution. By counting the gamma-radiation emitted by the plutonium it is possible to find the longitudinal distribution of the latter with an accuracy of better than' {+-} 2% in the case of perfect radial distribution; the last-named is usually determined by exposure on photographic emulsion. The axial distribution of the fuel density is determined with an accuracy of better than {+-} 1 Degree-Sign by measuring the non-absorbed fraction of gamma-radiation emitted by an external source, either Cs{sup 137} or lr{sup 192}. Cracks of around 10 {mu}m can be detected

  4. Non-destructive testing. V. 2

    International Nuclear Information System (INIS)

    Farley, J.M.; Nichols, R.W.

    1988-01-01

    The book entitled 'Non-destructive Testing' Volume 2, contains the proceedings of the fourth European Conference, organized by the British Institute of Non-Destructive Testing and held in London, September 1987. The volume contains seven chapters which examine the reliability of NDT, the economics of NDT and the use of NDT in:- civil engineering; oil, gas, coal and petrochemical industries; iron and steel industries; aerospace industry; and the nuclear and electricity supply industries. The seven chapters contain 78 papers, of which 19 are selected for INIS and indexed separately. (U.K.)

  5. Non-destructive testing of electronic parts

    International Nuclear Information System (INIS)

    Widenhorn, G.

    1980-01-01

    The requirements on quality, safety, faultlessness and reliability of electric components increase because of the high complexity of the appliances in which they are used. By means of examples a survey is given on the common non-destructive testing methods, testing operation and evaluation of test results on electric components which must meet in their application high requirements on quality and reliability. Defective components, especially those with hidden failures are sorted out by non-destructive testing and the failure frequency of the appliances and plants in testing and operation is greatly reduced. (orig.) [de

  6. European Non Destructive Examination Forum (ENDEF)

    International Nuclear Information System (INIS)

    Deffrennes, M.; Engl, G.; Estorff, U. von

    1998-01-01

    Non destructive examination (NDE) during fabrication, Pre-service inspection (PSI) and In service inspection (ISI) are considered key issues for the safe use of nuclear energy. They are important elements of plant lifetime management which is a critical item in decision making on nuclear policies. The European non destructive examination forum (ENDEF) founded by European Commission provides a platform for open discussion between representatives of the European industries with the purpose to establish cooperation between EU, Central and Eastern European Countries and New Independent States in the field of NDE and ISI

  7. Solution of a benchmark set problems for BWR and PWR reactors with UO2 and MOX fuels using CASMO-4

    International Nuclear Information System (INIS)

    Martinez F, M.A.; Valle G, E. del; Alonso V, G.

    2007-01-01

    In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO 2 ) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)

  8. Oxidative dissolution of unirradiated Mimas MOX fuel (U/Pu oxides) in carbonated water under oxic and anoxic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Odorowski, Mélina [CEA/DEN/DTCD/SECM/LMPA, BP 17171, 30207 Bagnols-sur-Cèze Cedex (France); MINES ParisTech, PSL Research University, Centre de Géosciences, 35 rue St Honoré, 77305 Fontainebleau (France); Jégou, Christophe, E-mail: christophe.jegou@cea.fr [CEA/DEN/DTCD/SECM/LMPA, BP 17171, 30207 Bagnols-sur-Cèze Cedex (France); De Windt, Laurent [MINES ParisTech, PSL Research University, Centre de Géosciences, 35 rue St Honoré, 77305 Fontainebleau (France); Broudic, Véronique; Peuget, Sylvain; Magnin, Magali; Tribet, Magaly [CEA/DEN/DTCD/SECM/LMPA, BP 17171, 30207 Bagnols-sur-Cèze Cedex (France); Martin, Christelle [Agence nationale pour la gestion des déchets radioactifs (Andra), DRD/CM, 1-7 rue Jean-Monnet, 92298 Châtenay-Malabry Cedex (France)

    2016-01-15

    Few studies exist concerning the alteration of Mimas Mixed-OXide (MOX) fuel, a mixed plutonium and uranium oxide, and data is needed to better understand its behavior under leaching, especially for radioactive waste disposal. In this study, two leaching experiments were conducted on unirradiated MOX fuel with a strong alpha activity (1.3 × 10{sup 9} Bq.g{sub MOX}{sup −1} reproducing the alpha activity of spent MOX fuel with a burnup of 47 GWd·t{sub HM}{sup −1} after 60 years of decay), one under air (oxic conditions) for 5 months and the other under argon (anoxic conditions with [O{sub 2}] < 1 ppm) for one year in carbonated water (10{sup −2} mol L{sup −1}). For each experiment, solution samples were taken over time and Eh and pH were monitored. The uranium in solution was assayed using a kinetic phosphorescence analyzer (KPA), plutonium and americium were analyzed by a radiochemical route, and H{sub 2}O{sub 2} generated by the water radiolysis was quantified by chemiluminescence. Surface characterizations were performed before and after leaching using Scanning Electron Microscopy (SEM), Electron Probe Microanalyzer (EPMA) and Raman spectroscopy. Solubility diagrams were calculated to support data discussion. The uranium releases from MOX pellets under both oxic and anoxic conditions were similar, demonstrating the predominant effect of alpha radiolysis on the oxidative dissolution of the pellets. The uranium released was found to be mostly in solution as carbonate species according to modeling, whereas the Am and Pu released were significantly sorbed or precipitated onto the TiO{sub 2} reactor. An intermediate fraction of Am (12%) was also present as colloids. SEM and EPMA results indicated a preferential dissolution of the UO{sub 2} matrix compared to the Pu-enriched agglomerates, and Raman spectroscopy showed the Pu-enriched agglomerates were slightly oxidized during leaching. Unlike Pu-enriched zones, the UO{sub 2} grains were much more

  9. KEOPS and other VENUS experiments dedicated to the criticality safety of a MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Lance, Benoit; Van Den Hende, Paul; Marloye, Daniel; Basselier, Jacques; Libon, Henri; De Vleeschhauwer, Marc; Moerenhout, Jeremie; Baeten, Peter

    2005-01-01

    The qualification scheme of criticality computer codes for Pu bearing powders lies upon databases which suffer from a lack of recent experimental results. As a MOX manufacture, BELGONUCLEAIRE is especially concerned by criticality safety and would like to address such an issue by launching with SCK-CEN an International Programme called KEOPS. (author)

  10. New tasks for non-destructive testing

    International Nuclear Information System (INIS)

    1990-01-01

    The proceedings contain 29 lectures and 43 posters which were presented in Trier at the annual meeting of the DGZfP in May 1990. The contributions report on further development of non-destructive testing methods towards more reliability, both of inspections and with regard to interpretation of the results. (MM) [de

  11. Utilization of radiation in non destructive tests

    International Nuclear Information System (INIS)

    Lopes, R.T.; Jesus, E.F.O. de; Junqueira, M.M.; Matos, J.A. de; Castello Branco, L.M.; Barros Junior, J.D.; Borges, J.C.

    1987-01-01

    The Nuclear Instrumentation Laboratory from COPPE/UFRJ has been developed techniques for using nuclear radiations to obtain images for non-destructive materials testing and medicine. With this objective, some prototypes of transmission computerized tomography systems using parallel beans and fan beans, with computer automation, including the mathematical process of image reprocessing and presentation in videos or printers are constructed [pt

  12. Burnup Credit of French PWR-MOx fuels: methodology and associated conservatisms with the JEFF-3.1.1 evaluation

    International Nuclear Information System (INIS)

    Chambon, A.

    2013-01-01

    Considering spent fuel management (storage, transport and reprocessing), the approach using 'fresh fuel assumption' in criticality-safety studies results in a significant conservatism in the calculated value of the system reactivity. The concept of Burnup Credit (BUC) consists in considering the reduction of the spent fuel reactivity due to its burnup. A careful BUC methodology, developed by CEA in association with AREVA-NC was recently validated and written up for PWR-UOx fuels. However, 22 of 58 French reactors use MOx fuel, so more and more irradiated MOx fuels have to be stored and transported. As a result, why industrial partners are interested in this concept is because taking into account this BUC concept would enable for example a load increase in several fuel cycle devices. Recent publications and discussions within the French BUC Working Group highlight the current interest of the BUC concept in PWR-MOx spent fuel industrial applications. In this case of PWR-MOx fuel, studies show in particular that the 15 FPs selected thanks to their properties (absorbing, stable, non-gaseous) are responsible for more than a half of the total reactivity credit and 80% of the FPs credit. That is why, in order to get a conservative and physically realistic value of the application k eff and meet the Upper Safety Limit constraint, calculation biases on these 15 FPs inventory and individual reactivity worth should be considered in a criticality-safety approach. In this context, thanks to an exhaustive literature study, PWR-MOx fuels particularities have been identified and by following a rigorous approach, a validated and physically representative BUC methodology, adapted to this type of fuel has been proposed, allowing to take fission products into account and to determine the biases related to considered isotopes inventory and to reactivity worth. This approach consists of the following studies: - isotopic correction factors determination to guarantee the criticality

  13. Economic importance of non-destructive testing

    International Nuclear Information System (INIS)

    Loebert, P.

    1979-01-01

    On May 21 to 23, 1979, the annual meeting of the Deutsche Gesellschaft fuer Zerstoerungsfreie Pruefung took place in Lindau near the Bodensee lake. About 600 experts from Germany and abroad participated in the meeting, whose general subject was 'The Economic Importance of Non-Destructive Testing'. Theoretical problems and practical investigations were discussed in a number of papers on special subjects. Apart from the 33 papers, there was also a poster show with 53 stands with texts, drawings, diagrams, and figures where the authors informed those interested on the latest state of knowledge in testing. The short papers were read in six sessions under the headings of rentability of non-destructive testing, X-ray methods, electromagnetic methods, and ultrasonic methods 1 and 2. (orig.) [de

  14. Non-destructive control of castings

    International Nuclear Information System (INIS)

    Boutault, J.; Mascre, C.

    1978-01-01

    The object of non-destructive control in foundries is to verify the metal structure, the absence of unacceptable discontinuity, total tightness, etc. This leads to a range of very varied controls according to the importance of the series, the quality level required by the specifications, the nature of the alloy. The originality of the solutions which are imperative for castings is shown through examples: casting of high quality complex forms in short series; very thick unit parts; very large series of parts requiring on efficient automation of non-destructive control. Lastly the publishing of testing methods and interpretating rules, which are the base of a friendly understanding between constructors and founders are recalled [fr

  15. Non-destructive control in nuclear construction

    International Nuclear Information System (INIS)

    Banus; Barbier; Launay

    1978-01-01

    Having recalled the characteristics of the fundamental components of the main primary circuit of nuclear boilers (900 MW) and the means appropriated for their control, it is recalled that the 'French Electricity Board's specifications and control rules' often prescribe more severe criteria than those existing in the U.S.A. Then practical examples of non-destructive controls concerning the steam generator end plates, vessel stainless steel linings, pump attachements, steam generator pipes are given [fr

  16. Development of non-destructive testing. Turkey

    International Nuclear Information System (INIS)

    1991-01-01

    A National Scheme for the qualification and certification of Non-Destructive Testing (NDT) personnel in various methods has been established as the first stage of implementation. Systematic training in such methods as radiography (RT), ultrasonics (UT), magnetic particles (MT), liquid penetrant (PT) and eddy currents (ET) at levels I, II and some at III has been initiated and should be continued. Direct link with the industry and continuous effort to extend practical applications is strongly recommended

  17. Safety performance comparation of MOX, nitride and metallic fuel based 25-100 MWe Pb-Bi cooled long life fast reactors without on-site refuelling

    International Nuclear Information System (INIS)

    Su'ud, Zaki

    2008-01-01

    In this paper the safety performance of 25-100 MWe Pb-Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb-Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance. The results of safety analysis of long life Pb-Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores. (author)

  18. MOX use in PWRs. EDF operation experience

    International Nuclear Information System (INIS)

    Provost, Jean-Luc; Debes, Michel

    2011-01-01

    From the origin, EDF back-end fuel cycle strategy has focused on 'closing the fuel cycle', in other words integrating fuel reprocessing, with vitrification of high level waste concentrated within small volumes, and the recycling of valuable materials. The implementation of this policy was marked in 1987 by the first loading of sixteen MOX. By December 2010, 20 reactors have been loaded with 1750 tHM of MOX. EDF current strategy is to match the reprocessing program with MOX manufacturing capacity to limit the quantity of separated plutonium. This is routinely called the 'flow ad-equation' strategy. Currently, the MOX Parity core management achieves balance of MOX and UOX performance with a significant increase of the MOX discharge burn-up. Globally, the behavior under irradiation of MOX fuel assemblies has been satisfactory. So far, from the beginning of MOX use in EDF PWRs, only 6 MOX FAs with rod leakage have been identified, which gives a very satisfactory level of reliability. The industrial maturity of MOX fuel, with increased performances, allows the improvement of nuclear KWh competitiveness and of the plant operation performance, while maintaining in operation the same safety level, without significant impact on environment and radiological protection. (author)

  19. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  20. The effect of dissolved hydrogen on the dissolution of 233U doped UO2(s) high burn-up spent fuel and MOX fuel

    International Nuclear Information System (INIS)

    Carbol, P.; Spahiu, K.

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of 233 U doped UO 2 (s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H 2 pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H 2 pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO 2 , high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10 -6 /yr - 10 -8 /yr with a recommended value of 4x10 -7 /yr for dissolved hydrogen concentrations above 10 -3 M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO 2 and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB

  1. A study on adsorption onto TODGA resin after electrolytic reduction in ERIX process for reprocessing spent FBR-MOX fuel

    International Nuclear Information System (INIS)

    Hoshi, Harutaka; Arai, Tsuyoshi; Wei, Yuezhou; Kumagai, Mikio; Asakura, Toshihide; Morita, Yasuji

    2005-01-01

    For reprocessing spent FBR-MOX fuel, an advanced aqueous reprocessing process ''ERIX process'' has been developed. In this system, hydrazine is used as reduction holding reagent for the valance adjustment of U by electrolytic reduction in nitric acid solution. Therefore, hydrazine is contained in high level liquid waste after separation of U, Pu and Np. Effect of hydrazine on adsorption of FP onto TODGA resin was examined. When hydrazine concentration was less than 0.3 M, effect on the distribution coefficient was negligibly small. After electrolytic reduction, some elements exist as lower valence state. Ru and Tc are most difficult elements to control their behavior in aqueous process. The distribution coefficient of both Ru and Tc onto TODGA decreased after electrolytic reduction, because they are reduced to lower valence. Hence, it is difficult for Ru or Tc to diffuse to allover the process and separation of MA from Tc and Ru was enhanced by electrolytic reduction. (author)

  2. Assessment of pin-by-pin fission rate distribution within MOX/UO{sub 2} fuel assembly using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Louis, Heba Kareem; Amin, Esmat [Nuclear and Radiological Regulation Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2016-03-15

    The aim of the present paper is to assess the calculations of pin-by-pin group integrated fission rates within MOX/UO{sub 2} Fuel assemblies using the Monte Carlo code MCNP2.7c with two sets of the available latest nuclear data libraries used for calculating MOX-fueled systems. The data that are used in this paper are based on the benchmark by the NEA Nuclear Science Committee (NSC). The k{sub ∞} and absorption/fission reaction rates per isotope, k{sub eff} and pin-by-pin group integrated fission rates on 1/8 fraction of the geometry are determined. To assess the overall pin-by-pin fission rate distribution, the collective per cent error measures were investigated. The results of AVG, MRE and RMS error measures were less than 1 % error. The present results are compared with other participants using other Monte Carlo codes and with CEA results that were taken in the benchmark as reference. The results with ENDF/B-VI.6 are close to the results received by MVP (JENDL3.2) and SCALE 4.2 (JEF2.2). The results with ENDF/BVII.1 give higher values of k{sub ∞} reflecting the changes in the newer evaluations. In almost all results presented here, the MCNP calculated results with ENDF/B VII.1 should be considered more than those obtained by using other Monte Carlo codes and nuclear data libraries. The present calculations may be consider a reference for evaluating the numerical schemes in production code systems, as well as the global performance including cross-section data reduction methods as the calculations used continuous energy and no geometrical approximations.

  3. Rough order of magnitude cost estimate for immobilization of 18.2 MT of plutonium sharing existing facilities at Hanford with MOX fuel fabrication facility: alternative 4B

    International Nuclear Information System (INIS)

    DiSabatino, A.

    1998-01-01

    The purpose of this Cost Estimate Report is to identify preliminary capital and operating costs for a facility to immobilize 18.2 metric tons (nominal) of plutonium as a ceramic in an existing facility at Hanford, the Fuels and Materials Examination Facility (FMEF). The MOX Fuel Fabrication Facility (MFFF), which is being costed in a separate report, will also be located in the FMEF in this co-location option

  4. Development of inspection data collection and evaluation system for large scale MOX fuel fabrication plant safeguards (3)

    International Nuclear Information System (INIS)

    Kumakura, Shinichi; Masuda, Shoichiro; Iso, Shoko; Hisamatsu, Yoshinori; Kurobe, Hiroko; Nakajima, Shinji

    2015-01-01

    Inspection Data Collection and Evaluation System is the system to store inspection data and operator declaration data collected from various measurement equipment, which is installed in fuel fabrication processes of the large-scale MOX fuel fabrication plant, and to make safeguards evaluation based on Near Real Time Accountancy (NRTA) using these data. Nuclear Material Control Center developed the simulator to simulate fuel fabrication process, in-process material inventory/flow data and the measurement data and the adequacy/impact to the uncertainty of the material balance using the simulation results, such as the facility operation and the operational status, has been reviewed. Following the 34th INMM Japan chapter presentation, the model similar to the real nuclear material accountancy during the fuel fabrication process was simulated and the nuclear material accountancy and its uncertainty (Sigma MUF) have been reviewed. Some findings have been obtained, such as regarding evaluation related indicators for verification under a more realistic accountancy which could be applied by operator. (author)

  5. Safeguards and Non-destructive Assay

    International Nuclear Information System (INIS)

    Carchon, R.; Bruggeman, M.

    2001-01-01

    SCK-CEN's programme on safeguards and non-destructive assay includes: (1) various activities to assure nuclear materials accountancy; (2) contributes to the implementation of Integrated Safeguards measures in Belgium and to assist the IAEA through the Belgian Support Programme; (3) renders services to internal and external customers in the field of safeguards; (4) improves passive neutron coincidence counting techniques for waste assay and safeguards verification measurements by R and D on correlation algorithms implemented via software or dedicated hardware; (5) improves gamma assay techniques for waste assay by implementing advanced scanning techniques and different correlation algorithms; and (6) develops numerical calibration techniques. Major achievements in these areas in 2000 are reported

  6. Non-destructive testing: significant facts

    International Nuclear Information System (INIS)

    Espejo, Hector; Ruch, Marta C.

    2006-01-01

    In the last fifty years different organisations, both public and private, have been assigned to the mission of introducing into the country the most relevant aspects of the modern technological discipline 'Non Destructive Testing' (NDT) through a manifold of activities, such as training and education, research, development, technical assistance and services, personnel qualification/certification and standardisation. A review is given of the significant facts in this process, in which the Argentine Atomic Energy Commission, CNEA, played a leading part, a balance of the accomplishments is made and a forecast of the future of the activity is sketched. (author) [es

  7. Ionizing radiations for non-destructive evaluation

    International Nuclear Information System (INIS)

    Raj, Baldev; Venkataraman, B.

    1989-01-01

    A state of the art of major non-destructive testing (NDT) techniques based on ionising radiations is presented. These techniques are broadly classified into three categories, namely, radiography, radiation gaging and analytical applications. The basic principles behind each method are explained and salient features of each technique which make it suitable for a particular task are described. Several illustrative applications drawn from the nuclear industry are given. The monograph is intended to serve as an introductory guide to scientist and engineers engaged in NDT activities. (M.G.B.). 32 refs., 13 figs., 5 tabs

  8. Non-Destructive Testing for Concrete Structure

    International Nuclear Information System (INIS)

    Tengku Sarah Tengku Amran; Noor Azreen Masenwat; Mohamad Pauzi Ismail

    2015-01-01

    Nondestructive testing (NDT) is a technique to determine the integrity of a material, component or structure. It is essential in the inspection of alteration, repair and new construction in the building industry. There are a number of non-destructive testing techniques that can be applied to determine the integrity of concrete in a completed structure. Each has its own advantages and limitations. For concrete, these problems relate to strength, cracking, dimensions, delamination, and inhomogeneities. NDT is reasonably good and reliable tool to measure the property of concrete which also gives the fair indication of the compressive strength development. This paper discussed the concrete inspection using combined methods of NDT. (author)

  9. Performance of MOX fuel: An overview of the experimental programme of the OECD Halden Reactor Project and review of selected results

    International Nuclear Information System (INIS)

    Wiesenack, W.; McGrath, M.

    2000-01-01

    The OECD Halden Reactor Project has defined an extensive experimental programme related to MOX fuels which is being executed with the objective to provide a performance data base similar to that available for UO 2 . In addition to utilising fresh MOX fuel and re-instrumented segments from LWR irradiations to high burnup, the concept of inert matrix fuel is being addressed. The irradiation in the Halden reactor is performed in rigs allowing steady state, power ramping and cyclic operation. In-pile data are obtained from instrumentation such as fuel centreline thermocouples, pressure transducers, fuel and cladding elongation detectors, and movable gauges for measuring the diametral deformation. Various phenomena can be assessed in this way, e.g. thermal performance, swelling and densification, PCMI and fission gas release. The paper describes the objectives of various experiments and provides examples of temperature, pressure and cladding elongation measurements performed on MOX fuel. Salient results are related to the threshold for the onset of significant fission gas release and the relaxation behaviour in a power ramp-PCMI situation. (author)

  10. Fission gas release behavior of MOX fuels under simulated daily-load-follow operation condition. IFA-554/555 test evaluation with FASTGRASS code

    International Nuclear Information System (INIS)

    Ikusawa, Yoshihisa; Ozawa, Takayuki

    2008-03-01

    IFA-554/555 load-follow tests were performed in HALDEN reactor (HBWR) to study the MOX fuel behavior under the daily-load-follow operation condition in the framework of ATR-MOX fuel development in JAEA. IFA-554/555 rig had the instruments of rod inner pressure, fuel center temperature, fuel stack elongation, and cladding elongation. Although the daily-load-follow operation in nuclear power plant is one of the available options for economical improvement, the power change in a short period in this operation causes the change of thermal and mechanical irradiation conditions. In this report, FP gas release behavior of MOX fuel rod was evaluated under the daily-load-follow operation condition with the examination data from IFA-554/555 by using the computation code 'FASTGRASS'. From the computation results of FASTGRASS code which could compute the FP gas release behavior under the transient condition, it could be concluded that FP gas was released due to the relaxation of fuel pellet inner stress and pellet temperature increase, which were caused by the cyclic power change during the daily-load-follow operation. In addition, since the amount of released FP gas decreased during the steady operation after the daily-load-follow, it could be mentioned that the total of FP gas release at the end of life with the daily-load-follow is not so much different from that without the daily-load-follow. (author)

  11. Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

    International Nuclear Information System (INIS)

    Ludewigt, Bernhard; Mozin, Vladimir; Campbell, Luke; Favalli, Andrea; Hunt, Alan W.; Reedy, Edward T.E.; Seipel, Heather

    2015-01-01

    High-energy, beta-delayed gamma-ray spectroscopy is a potential, non-destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried out at the IAC using a photo-neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-energy delayed gamma rays from 235 U, 239 Pu, and 241 Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241 Pu, a significant fissile constituent in spent fuel, was measured and compared to 239 Pu. The 241 Pu/ 239 Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-ray emission was developed and demonstrated on a limited 235 U data set. De-convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-rate LaBr 3 detectors was investigated as a potential alternative to HPGe detectors. Modeling capabilities were added to an

  12. Confirmation test of powder mixing process in J-MOX

    International Nuclear Information System (INIS)

    Ota, Hiroshi; Osaka, Shuichi; Kurita, Ichiro

    2009-01-01

    Japan Nuclear Fuel Ltd. (hereafter, JNFL) MOX Fuel Fabrication Plant (hereafter, J-MOX) is what fabricates MOX fuel for domestic light water power plants. Development of design concept of J-MOX was started mid 90's and the frame of J-MOX process was clarified around 2000 including adoption of MIMAS process as apart of J-MOX powder process. JNFL requires to take an answer to any technical question that has not been clarified ever before by world's MOX and/or Uranium fabricators before it commissions equipment procurement. J-MOX is to be constructed adjacent to the Rokkasho Reprocessing Plant (RRP) and to utilize MH-MOX powder recovered at RRP. The combination of the MIMAS process and the MH-MOX powder is what has never tried in the world. Therefore JNFL started a series of confirmation tests of which the most important is the powder test to confirm the applicability of MH-MOX powder to the MIMAS process. The MH-MOX powder, consisting of 50% plutonium oxide and 50% uranium oxide, originates JAEA development utilizing microwave heating (MH) technology. The powder test started with laboratory scale small equipment utilizing both uranium and the MOX powder in 2000, left a solution to tough problem such as powder adhesion onto equipment, and then was followed by a large scale equipment test again with uranium and the MOX powder. For the MOX test, actual size equipment within glovebox was manufactured and installed in JAEA plutonium fuel center in 2005, and based on results taken so far an understanding that the MIMAS equipment, with the MH-MOX powder, can present almost same quality MOX pellet as what is introduced as fabricated in Europe was developed. The test was finished at the end of Japanese fiscal year (JFY) 2007, and it was confirmed that the MOX pellets fabricated in this test were almost satisfied with the targeted specifications set for domestic LWR MOX fuels. (author)

  13. Non-destructive testing; Examenes no destructivos

    Energy Technology Data Exchange (ETDEWEB)

    Calva, Mauricio; Loske, Achim [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1988-12-31

    The application of non-destructive testing (NDT) in several technical and industrial fields is pointed out, standing out its utilization in the detection of future failures without affecting the examined element. Likewise, the different types of NDTs and their processes, such as x-rays, ultrasoud, magnetic particles, induced currents, penetrating fluids, and optical means, are described. The Non-Destructive Tests Laboratory of the Instituto de Investigaciones Electricas (IIE), plans to create new and more reliable systems independent from the operator`s capacity, to contribute to fulfill the inspection and quality control needs of the generating Mexican power plants. [Espanol] Se senala la aplicacion de los examenes no destructivos (END) a diversos campos tecnicos e industriales, destacando su utilizacion en la deteccion de futuras fallas sin afectar el elemento examinado. Asimismo, se describen los diferentes tipos de END y sus procesos, tales como radiografia, ultrasonido, particulas magneticas, corrientes inducidas, liquidos penetrantes y metodos opticos. El Laboratorio de Pruebas no Destructivas, del Instituto de Investigaciones Electricas (IIE), planea crear sistemas novedosos mas confiables, que no dependan de la capacidad del operador, para contribuir a satisfacer las necesidades de inspeccion y control de calidad que se presentan en las plantas generadoras de energia mexicanas.

  14. Developments in non-destructive beam diagnostics

    International Nuclear Information System (INIS)

    Fraser, J.S.

    1981-01-01

    With the large average beam currents being achieved in accelerators and storage rings, there is an increasing need for non-destructive beam diagnostic devices. For continuous beams, position monitors of the capacitive pick-up type are replaced by resonant devices that respond to the transverse displacement of the beam centroid. Bunch length monitors of the SLAC type using resonant cavities operating in the TM 010 mode can be used for continuous beams. The more detailed information derivable from beam profile scanners requires development of improved non-destructive devices. Profile monitors which scan the visible light produced by high current beams may be more reliable than ones using the residual ionization if the light intensity from gas molecules following nonionizing collisions with beam particles gives a measure of the beam current density independent of the local electron density. The intense Balmer series lines from neutral hydrogen beams have been used successfully to measure beam profiles. At CRNL and at LASL, beam light profile monitors are being developed for high average current accelerators. Three or more projections will be recorded to allow tomographic reconstruction of the two-dimensional beam current density. Light detection is either by intensified Reticons or ISIT vidicons. The use of three or more beam light monitors on a beam transport line will also permit estimates of the transverse emittance to be made through the reconstruction technique

  15. Non-destructive testing; Examenes no destructivos

    Energy Technology Data Exchange (ETDEWEB)

    Calva, Mauricio; Loske, Achim [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1987-12-31

    The application of non-destructive testing (NDT) in several technical and industrial fields is pointed out, standing out its utilization in the detection of future failures without affecting the examined element. Likewise, the different types of NDTs and their processes, such as x-rays, ultrasoud, magnetic particles, induced currents, penetrating fluids, and optical means, are described. The Non-Destructive Tests Laboratory of the Instituto de Investigaciones Electricas (IIE), plans to create new and more reliable systems independent from the operator`s capacity, to contribute to fulfill the inspection and quality control needs of the generating Mexican power plants. [Espanol] Se senala la aplicacion de los examenes no destructivos (END) a diversos campos tecnicos e industriales, destacando su utilizacion en la deteccion de futuras fallas sin afectar el elemento examinado. Asimismo, se describen los diferentes tipos de END y sus procesos, tales como radiografia, ultrasonido, particulas magneticas, corrientes inducidas, liquidos penetrantes y metodos opticos. El Laboratorio de Pruebas no Destructivas, del Instituto de Investigaciones Electricas (IIE), planea crear sistemas novedosos mas confiables, que no dependan de la capacidad del operador, para contribuir a satisfacer las necesidades de inspeccion y control de calidad que se presentan en las plantas generadoras de energia mexicanas.

  16. Potential development of non-destructive assay for nuclear safeguards

    International Nuclear Information System (INIS)

    Benoit, R.; Cuypers, M.; Guardini, S.

    1983-01-01

    After a brief summary on the role of non-destructive assay in safeguarding the nuclear fuel cycle, its evolution from NDA methods development to other areas is illustrated. These areas are essentially: a) the evaluation of the performances of NDA techniques in field conditions; b) introduction of full automation of measurement instrument operation, using interactive microprocessors and of measurement data handling evaluation and retrieval features; c) introduction of the adequate link and compatibility to assure NDA measurement data transfer in an integrated safeguards data evaluation scheme. In this field, the Joint Research Centre (JRC) of the Commission of the European Communities (CEC) is developing and implementing a number of techniques and methodologies allowing an integrated and rational treatment of the large amount of safeguards data produced. In particular for the non-destructive assay measurements and techniques, the JRC has studied and tested methodologies for the automatic generation and validation of data of inventory verification. In order to apply these techniques successfully in field, the JRC has studied the design requirements of NDA data management and evaluation systems. This paper also discusses the functional requirements of an integrated system for NDA safeguards data evaluation

  17. Non-destructive nuclear forensics of radioactive samples

    Energy Technology Data Exchange (ETDEWEB)

    Rogge, R.B. [Canadian Neutron Beam Centre, Chalk River, ON (Canada); Alexander, Q.; Bentoumi, G.; Dimayuga, F. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Flacau, R. [Canadian Neutron Beam Centre, Chalk River, ON (Canada); Li, G.; Li, L.; Sur, B. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    It is a matter of public safety and security to be able to examine suspicious packages of unknown origin. If the package is radioactive and sealed (i.e., the radioactive materials contained in the package, including their chemical and physical forms, are unknown), there is a significant risk on how to handle the package and eventually safely dispose of its contents. Within the context of nuclear security, nuclear forensics helps address the key issue of identifying the nature and origin of radioactive and nuclear material in order to improve physical protection measures and prevent future theft or diversion of these materials. Nuclear forensics utilizes analytical techniques, destructive and non-destructive, developed for applications related to nuclear fuel cycles. This paper demonstrates the non-destructive examination techniques that can be used to inspect encapsulated radioactive samples. Results of γ spectroscopy, X-ray spectroscopy, neutron imaging, neutron diffraction, and delayed neutron analysis as applied to an examination of sealed capsules containing unknown radioactive materials are presented. The paper also highlights the value of these techniques to the overall nuclear forensic investigation to determine the origin of these unknown radioactive materials. (author)

  18. MOx Depletion Calculation Benchmark

    International Nuclear Information System (INIS)

    San Felice, Laurence; Eschbach, Romain; Dewi Syarifah, Ratna; Maryam, Seif-Eddine; Hesketh, Kevin

    2016-01-01

    Under the auspices of the NEA Nuclear Science Committee (NSC), the Working Party on Scientific Issues of Reactor Systems (WPRS) has been established to study the reactor physics, fuel performance, radiation transport and shielding, and the uncertainties associated with modelling of these phenomena in present and future nuclear power systems. The WPRS has different expert groups to cover a wide range of scientific issues in these fields. The Expert Group on Reactor Physics and Advanced Nuclear Systems (EGRPANS) was created in 2011 to perform specific tasks associated with reactor physics aspects of present and future nuclear power systems. EGRPANS provides expert advice to the WPRS and the nuclear community on the development needs (data and methods, validation experiments, scenario studies) for different reactor systems and also provides specific technical information regarding: core reactivity characteristics, including fuel depletion effects; core power/flux distributions; Core dynamics and reactivity control. In 2013 EGRPANS published a report that investigated fuel depletion effects in a Pressurised Water Reactor (PWR). This was entitled 'International Comparison of a Depletion Calculation Benchmark on Fuel Cycle Issues' NEA/NSC/DOC(2013) that documented a benchmark exercise for UO 2 fuel rods. This report documents a complementary benchmark exercise that focused on PuO 2 /UO 2 Mixed Oxide (MOX) fuel rods. The results are especially relevant to the back-end of the fuel cycle, including irradiated fuel transport, reprocessing, interim storage and waste repository. Saint-Laurent B1 (SLB1) was the first French reactor to use MOx assemblies. SLB1 is a 900 MWe PWR, with 30% MOx fuel loading. The standard MOx assemblies, used in Saint-Laurent B1 reactor, include three zones with different plutonium enrichments, high Pu content (5.64%) in the center zone, medium Pu content (4.42%) in the intermediate zone and low Pu content (2.91%) in the peripheral zone

  19. Integrated software package for nuclear material safeguards in a MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Schreiber, H.J.; Piana, M.; Moussalli, G.; Saukkonen, H.

    2000-01-01

    Since computerized data processing was introduced to Safeguards at large bulk handling facilities, a large number of individual software applications have been developed for nuclear material Safeguards implementation. Facility inventory and flow data are provided in computerized format for performing stratification, sample size calculation and selection of samples for destructive and non-destructive assay. Data is collected from nuclear measurement systems running in attended, unattended mode and more recently from remote monitoring systems controlled. Data sets from various sources have to be evaluated for Safeguards purposes, such as raw data, processed data and conclusions drawn from data evaluation results. They are reported in computerized format at the International Atomic Energy Agency headquarters and feedback from the Agency's mainframe computer system is used to prepare and support Safeguards inspection activities. The integration of all such data originating from various sources cannot be ensured without the existence of a common data format and a database system. This paper describes the fundamental relations between data streams, individual data processing tools, data evaluation results and requirements for an integrated software solution to facilitate nuclear material Safeguards at a bulk handling facility. The paper also explains the basis for designing a software package to manage data streams from various data sources and for incorporating diverse data processing tools that until now have been used independently from each other and under different computer operating systems. (author)

  20. Non destructive testing in amusement park

    International Nuclear Information System (INIS)

    Dominguez Marrero, Humberto; Hernandez Torres, Debora; Sendoya Puente, Felix; Herrera Palma, Victoria; Suarez Guerra, Yarelis; Moreno Hernandez, Eduardo; Lopez Hernandez, Pedro

    2009-01-01

    In 2006 began the installation of Chinese amusement parks at several places in Havana City. Structural security is one of the principal tasks that should be done, since the beginning of the services of these installations. The use on Non Destructive Testing Techniques (NDT), has to be development and implemented in order to avoid the possibility of failure during services with a consequence threat to safety for the public presented. In this work it is shown the results of application of NDT techniques and recommendations for the quality control of the different welds and mechanical components presented. Techniques as Visual Examination, Liquid Penetrant and Ultrasonic have been used for these purposes in order to obtain a structural diagnostic in the amusement parks. There are also exposed the use and implementation of international recommendations and Standards, which are very rigorous in its applications for the case of recreation industry. This is a consequence to its social service fundamentally to children and teenage people. (Author)

  1. Development of automation and remotisation systems for fabrication of (Th-233U)O2 MOX fuel for AHWR

    International Nuclear Information System (INIS)

    Saraswat, Anupam; Danny, K.M.; Chakraborty, S.; Somayajulu, P.S.; Kumar, Arun; Mittal, R.; Prasad, R.S.; Mahule, K.N.; Panda, S.; Jayarajan, K.

    2011-01-01

    To meet the ever increasing power requirement of India, country is planning to utilize its large thorium reserves for the third stage of nuclear power program based on Thorium-Uranium 233 fuel in A.H.W.R. Although there are many advantages of (Th- 233 U)O 2 fuel cycle, presence of radiological hazards due to the presence of 1000-2000 ppm level of 232 U in the 233 U fuel and inertness of ThO 2 makes handling and fabrication of fuel difficult. The associated high alpha and gamma activity demands high level of automation and remote handling in alpha tight hot cells. To demonstrate automation and remotisation in (Th- 233 U)O 2 fuel fabrication, a mock up facility is being set up at BARC. This facility shall develop automation systems required for remote fuel fabrication in a simulated hot cell environment. There are many innovative schemes and systems being developed like integrated powder pellet system, remote viewing system for hot cell application etc. Low visibility inside the hot cell has always been a problem for the operator. To overcome this problem a remote viewing system has been developed by which entire hot cell area can be scanned with the use of a joystick and the display can be seen on a LCD monitor. The viewing system is made up of radiation resistant optics which can work even in high gamma fields. It consists of objective end assembly which is used to scan the hot cell area with the help of prism doublets and drive mechanism for capturing full 360 deg solid angle view. There is a Galilean telescope and focusing system used for focusing images of distant objects. Drive mechanism can be controlled by the joystick available to the operator. System has a high resolution CCD display and camera which gives a clear display of objects lying inside the hot cell area. Integrated powder pellet system is being developed for fabrication of MOX pellets from feed powder. This will be automated system which will take input in the form of MOX powder and convert it

  2. A utility analysis of MOX recycling policy

    International Nuclear Information System (INIS)

    Pfaeffli, J.L.

    1990-01-01

    The author presents the advantages of recycling of plutonium and uranium from spent reactor fuel assemblies as follows: natural uranium and enrichment savings, mixed oxide fuel (MOX) fuel assembly cost, MOX compatibility with plant operation, high burnups, spent MOX reprocessing, and non-proliferation aspects.Disadvantages of the recycling effort are noted as well: plutonium degradation with time, plutonium availability, in-core fuel management, administrative authorizations by the licensings authorities, US prior consent, and MOX fuel fabrication capacity. Putting the advantages and disadvantages in perspective, it is concluded that the recycling of MOX in light water reactors represents, under the current circumstances, the most appropriate way of making use of the available plutonium

  3. The nuclear future; prospects for reprocessing and mixed oxide nuclear fuel; why use MOX in civil reactors?

    International Nuclear Information System (INIS)

    Bay, H.

    2002-01-01

    There are many answer to the question 'Why use MOX in civil reactors?'. The most likely one is because plutonium is an energy source and MOX is used when it is economic to do so. Other incentives include the reduction of global separated plutonium stocks and the subsequent potential reduction of proliferation risk. (author)

  4. A Windows-95 based DAS software for the determination of chloride and fluoride in MOX fuel samples by pyrohydrolysis-ISE method

    Energy Technology Data Exchange (ETDEWEB)

    Govindan, R; Kulkarni, Atul; Sivashankaran, G; Raul, Seema; Behere, P G; Mallik, G K; Bhargava, V K; Kamath, H S [Advanced Fuel Fabrication Facility, Bhabha Atomic Research Centre, Tarapur (India)

    1999-01-01

    A Windows-95 based Data Acquisition System (DAS) software has been developed to determine chloride and fluoride in MOX (Mixed Oxide) fuel pellets by pyrohydrolysis followed by Ion Selective Electrodes (ISE). The DAS software controls the furnace remotely and acquires data related to ion-85 ion analyser through RS-232C serial port. This results in ease of operation of furnace during pyrohydrolysis and storage of measured data from ion-85 ion analyser for future use. (author) 4 refs., 1 fig., 3 tabs.

  5. Control of nuclear material hold-up: The key factors for design and operation of MOX fuel fabrication plants in Europe

    International Nuclear Information System (INIS)

    Beaman, M.; Beckers, J.; Boella, M.

    2001-01-01

    Full text: Some protagonists of the nuclear industry suggest that MOX fuel fabrication plants are awash with nuclear materials which cannot be adequately safeguarded and that materials 'stuck in the plant' could conceal clandestine diversion of plutonium. In Europe the real situation is quite different: nuclear operators have gone to considerable efforts to deploy effective systems for safety, security, quality and nuclear materials control and accountancy which provide detailed information. The safeguards authorities use this information as part of the safeguards measures enabling them to give safeguards assurances for MOX fuel fabrication plants. This paper focuses on the issue of hold-up: definition of the hold-up and of the so-called 'hidden inventory'; measures implemented by the plant operators, from design to day to day operations, for minimising hold-up and 'hidden inventory'; plant operators' actions to manage the hold-up during production activities but also at PIT/PIV time; monitoring and management of the 'hidden inventory'; measures implemented by the safeguards authorities and inspectorate for verification and control of both hold-up and 'hidden inventory'. The examples of the different plant specific experiences related in this paper reveal the extensive experience gained in european MOX fuel fabrication plants by the plant operators and the safeguards authorities for the minimising and the control of both hold-up and 'hidden inventory'. MOX fuel has been fabricated in Europe, with an actual combined capacity of 2501. HM/year subject, without any discrimination, to EURATOM Safeguards, for more than 30 years and the total output is, to date, some 1000 t.HM. (author)

  6. Mox pellet reference material

    International Nuclear Information System (INIS)

    Perolat, J.P.

    1991-01-01

    A first batch of MOX pellets certified in plutonium and uranium has been prepared and characterised in France to meet the needs of laboratories which are engaged upon destructive analysis for safeguards purposes especially in fuel fabrication plants. The pellets sintering has been obtained in a special fabrication to achieve an homogeneity better than 0.1%. The plutonium and uranium characterisation by chemical analysis has been carried out by two laboratories using at least two different methods. 1 fig., 5 refs

  7. 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

  8. Efficiency evaluation test of waste non-destructive analysis device

    International Nuclear Information System (INIS)

    Maeda, Kouichi; Ogasawara, Kensuke; Nisizawa, Ichio

    2000-03-01

    A device for non-destructive analysis of plutonium in alpha solid waste has been installed in NUCEF; Nuclear Fuel Cycle Safety Engineering Research Facility. The device has been designed to determine the amount of radioisotopes in carton-boxes, 45 l steel cans and 200 l steel cans containing relatively low density waste. Considering the waste density and the heterogeneity of radio-sources, the proper distance between the detector and the waste, and the open degree of the collimator have been settled, because real waste may contain several kinds of material and the heterogeneity of radioactivity. It has been confirmed from the evaluation of the detect limit that plutonium of about 8 MBq can be determined with the accuracy of 10% and the device may be proper for the practical application. (author)

  9. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P. [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K. (ed.) [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  10. Development of vibropac MOX fuel pins serviceable up TP superhigh burnups

    International Nuclear Information System (INIS)

    Mayorshin, A.A.; Gadzhiev, G.I.; Kisly, V.A.; Skiba, O.V.; Tzykanov, V.A.

    1998-01-01

    The main results on investigations of fast reactor fuel pins with (UPu)O 2 vibropac fuel to substantiate their serviceability up to the super-high burnups are presented. The BOR-60 reactor fuel pins radiation behaviour in stationary, transient and designed emergency conditions has been determined from the fuel pins dimensional stability analysis having regard to the results of investigation fuel and cladding swelling as well as estimations of fuel and cladding thermal-mechanical and physico-chemical interactions. It is shown that the change of the outer diameter is minimum in fuel pins with VMOX fuel with a getter-metallic uranium powder and ferrito-martensite steel cladding, and the corrosion damage of the cladding inner surface is absent up to 26% h.a. The experiments with over-heating of the irradiated fuel pins cladding up to 850 deg. C did not lead to any changes in pins integrity. The availability of the periphery area of the vibropac fuel cure initial structure provides the minimum level of the thermal-mechanical stress at transient conditions of reactor operation. (author)

  11. The United States pit disassembly and conversion project -- Meeting the MOX fuel specification

    International Nuclear Information System (INIS)

    Nelson, T.O.; James, C.A.; Kolman, D.G.

    1998-01-01

    The US is actively involved in demonstrating the disassembly of nuclear weapons pits to an unclassified form readied for disposition. The MOX option is the most likely path forward for plutonium that originated from nuclear weapon pits. The US demonstration line for pit disassembly and conversion is known as ARIES, the advanced recovery and integrated extraction system. The ARIES demonstration line is being used to gather data in an integrated fashion of the technologies needed for pit disassembly and conversion. These activities include the following modules: pit bisection, hydride-dehydride, oxide conversion, canning, electrolytic decontamination, and nondestructive assay (NDA). Pit bisection swages in a pit in half. Hydride-dehydride converts the pit plutonium metal to an unclassified metal button. To convert the plutonium metal to an oxide the US is investigating a number of options. The primary oxide conversion approach involves variations of combining plutonium hydriding and subsequent oxidation. Another approach is to simply oxidize the metal under controlled conditions-direct metal oxidation (DMO). To remove the gallium from the plutonium oxide, a thermal distillation approach is being used. These pyrochemical approaches will substantially reduce the wastes produced for oxide conversion of weapon plutonium, compared to traditional aqueous processing. The packaging of either the plutonium metal or oxide to long term storage criteria involves the canning and electrolytic decontamination modules. The NDA suite of instruments is then used to assay the material in the containers, which enables international verification without the need to open the containers and repackage them. All of these processes are described

  12. Current Status of J-MOX Safeguards Design and Future Prospects

    International Nuclear Information System (INIS)

    Sampei, T.; Hiruta, K.; Shimizu, J.; Ikegame, K.

    2015-01-01

    The construction of JNFL MOX Fuel Fabrication Plant (J-MOX) is proceeding toward active test using uranium and MOX in July 2017, and completion of construction in October 2017. Although the construction schedule is largely impacted by progress of licencing, according to domestic law, JNFL is making every effort to get necessary permission of business licence and authorization of design and construction method as soon as possible. On the other hand, it is desirable that integrated safeguards approach is effective, efficient and consistent with J-MOX facility features. Discussion about the approach is going on among IAEA, Japan Safeguards Office (JSGO) and JNFL, and IAEA is planning to introduce the measures into the approach such as application of Near Real-Time Accountancy with frequent declaration from operator, Containment/Surveillance measures to storages, internal flow verification with 100%, random interim inspection (RII) and so on. RII scheme is intended to increase efficiency without compromising effectiveness and makes interruption of facility operation minimum. Also newly developed and improved safeguards equipment will be employed and it is possible to realize to increase credibility and efficiency of inspection by introduction of unattended/automatic safeguards equipment. Especially IAEA and JSGO share the development of non-destructive assay systems which meet the requirements from both parties. These systems will be jointly utilized at the flow verification, RII and PIV. JNFL will continue to provide enough design information in a timely manner toward early establishment of safeguards approach for J-MOX. Also JNFL will implement the coordination of installation and commissioning of safeguards equipment, and Design Information Verification activities for completion of construction in October 2017

  13. Uranium, Plutonium and Neptunium Co-recovery with Irradiated Fast Reactor MOX Fuel by Single Cycle Extraction Process

    Energy Technology Data Exchange (ETDEWEB)

    Masaumi Nakahara; Yuichi Sano; Kazunori Nomura; Tadahiro Washiya; Jun Komaki [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1194 (Japan)

    2008-07-01

    The behavior of Np in single cycle extraction processes using tri-n-butylphosphate (TBP) as an extractant for U, Pu and Np co-recovery was investigated as a part of NEXT (New Extraction System for Transuranium) process. Two approaches for Np co-recovery with U and Pu were carried out with irradiated MOX fuel from fast reactor 'JOYO'; one was the counter current experiment using a feed solution with a high HNO{sub 3} concentration and the other used a scrubbing solution with a high HNO{sub 3} concentration. Experimental results showed that the leakage of Np to the raffinate were 0.986 % and 5.96 % under the condition of high HNO{sub 3} concentration in the feed solution and scrubbing solution, respectively. The simulation results based on these experiments indicated that most of Np could be extracted and co-recovered with U and Pu, just by increasing HNO{sub 3} concentrations in the feed and scrubbing solution on the single cycle extraction process. (authors)

  14. European Non Destructive Examination Forum (ENDEF)

    Energy Technology Data Exchange (ETDEWEB)

    Deffrennes, M [EC, DG XVII, Nuclear Energy, Brussels (Belgium); Engl, G [Siemens AG Energieerzeugung KWU, Erlangen (Germany); Estorff, U von [EC, JRC/IAM, Petten (Netherlands)

    1998-11-01

    ENDEF, an initiative of the European Commission, DG XVII (Energy) was well supported by the European industrial institutions working in assistance with nuclear industrial organisations in the CEEC`s (Central and Eastern European Countries) and NIS`s (New Independent States). This Forum provides effectively a platform for open discussion between representatives of industrial actors active in the NDE (Non Destructive Examination)/ISI (In Service Inspection) field with the purpose to establish a co-operation pattern between qualified representatives of the EU (European Union) industry to offer a better co-ordinated and well defined assistance to the CEEC`s and NIS in the field of NDE/ISI, and to lay the ground for further industrial co-operation. ENDEF developed a strategy to follow for the establishment of co-operation projects. This strategy is now used to understand the extent of past or present assistance projects and to identify the areas where more co-operation is needed. ENDEF encourages the creation in the NIS`s and CEEC`s of similar forums in order to increase the co-operation and co-ordination. ENDEF is also working in perfect agreement with the European Network ENIQ, piloted by the European plant operators. This identity of views lead to the leadership by the ENDEF co-ordinator of the ENIQ Task 3 involving Applications of the European Methodology for ISI qualification in the CEEC`s and NIS`s and presently fully integrated in ENDEF. (orig.)

  15. European Non Destructive Examination Forum (ENDEF)

    International Nuclear Information System (INIS)

    Deffrennes, M.; Engl, G.; Estorff, U. von

    1998-01-01

    ENDEF, an initiative of the European Commission, DG XVII (Energy) was well supported by the European industrial institutions working in assistance with nuclear industrial organisations in the CEEC's (Central and Eastern European Countries) and NIS's (New Independent States). This Forum provides effectively a platform for open discussion between representatives of industrial actors active in the NDE (Non Destructive Examination)/ISI (In Service Inspection) field with the purpose to establish a co-operation pattern between qualified representatives of the EU (European Union) industry to offer a better co-ordinated and well defined assistance to the CEEC's and NIS in the field of NDE/ISI, and to lay the ground for further industrial co-operation. ENDEF developed a strategy to follow for the establishment of co-operation projects. This strategy is now used to understand the extent of past or present assistance projects and to identify the areas where more co-operation is needed. ENDEF encourages the creation in the NIS's and CEEC's of similar forums in order to increase the co-operation and co-ordination. ENDEF is also working in perfect agreement with the European Network ENIQ, piloted by the European plant operators. This identity of views lead to the leadership by the ENDEF co-ordinator of the ENIQ Task 3 involving Applications of the European Methodology for ISI qualification in the CEEC's and NIS's and presently fully integrated in ENDEF. (orig.)

  16. LTA Physics Design: Description of All MOX Pin LTA Design

    International Nuclear Information System (INIS)

    Pavlovichev, A.M.

    2001-01-01

    In this document issued according to Work Release 02.P.99-1b the results of neutronics studies of > MOX LTA design are presented. The parametric studies of infinite MOX-UOX grids, MOX-UOX core fragments and of VVER-1000 core with 3 MOX LTAs are performed. The neutronics parameters of MOX fueled core have been performed for the chosen design MOX LTA using the Russian 3D code BIPR-7A and 2D code PERMAK-A with the constants prepared by the cell spectrum code TVS-M

  17. The data acquisition system for the management of nuclear materials involved in the fabrication of MOX fuel at the Cogema plant in Cadarache

    International Nuclear Information System (INIS)

    Crousilles, M.; Beche, M.; Dalverny, G.

    2001-01-01

    This article presents the follow-up system of all the nuclear materials that are involved in the industrial process of MOX fuel fabrication. This system, called Concerto, allows the management of MOX fabrication but also of any nuclear material transfer and of the stockpile of nuclear materials with taking into account their own specificity such as the risk of criticality. Operators that intervene on the different steps of the fabrication process, supply Concerto with information so Concerto can be considered as a near real-time system providing and recording the localization, the composition, the weight, the container,... of any batch of nuclear materials. Concerto complies with the requirements of quality assurance but also of nuclear safety by forbidding any transfer whenever the maximal authorized quantity would be exceeded. (A.C.)

  18. 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.

  19. MOX in reactors: present and future

    International Nuclear Information System (INIS)

    Arslan, Marc; Gros, Jean Pierre; Niquille, Aurelie; Marincic, Alexis

    2010-01-01

    In Europe, MOX fuel has been supplied by AREVA for more than 30 years, to 36 reactors: 21 in France, 10 in Germany, 3 in Switzerland, 2 in Belgium. For the present and future, recycling is compulsory in the frame of sustainable development of nuclear energy. By 2030 the overall volume of used fuel will reach about 400 000 t worldwide. Their plutonium and uranium content represents a huge resource of energy to recycle. That is the reason why, the European Utilities issued an EUR (European Utilities Requirement) demanding new builds reactors to be able of using MOX Fuel Assemblies in up to 50 % of the core. AREVA GEN3+ reactors, like EPR TM or ATMEA TM designed with MHI partnership, are designed to answer any utility need of MOX recycling. The example of the EPR TM reactor operated with 100 % MOX core optimized for MOX recycling will be presented. A standard EPR TM can be operated with 100 % MOX core using an advanced homogeneous MOX (single Pu content) with highly improved performances (burn-up and Cycle length). The adaptations needed and the main operating and safety reactor features will be presented. AREVA offers the utilities throughout the world, fuel supply (UO 2 , ERU, MOX), and reactors designed with all the needed capability for recycling. For each country and each utility, an adapted global solution, competitive and non proliferant can be proposed. (authors)

  20. Combining data in non-destructive testing

    International Nuclear Information System (INIS)

    Lavayssiere, B.

    1994-03-01

    Non-destructive testing of some components requires quite often the use of several methods such as X-ray, ultrasonics, Eddy Currents. But the efficiency of a NDT method is highly dependent on the fact that the detectability of flaws in a specimen relies on the choice of the best method. Moreover a lot of inspection issues could benefit from the use of more than one test method, as each NDT method has its own physical properties and technological limits. Some questions still remain: how to combine data, at what level and for what functionality. Simple monomethod processes are well-known now. They include techniques like reconstruction which belongs to the so-called ill-posed problems in the field of mathematics. For NDT data processing, it has the ability to estimate real data from distorted ones coming from a probe. But, up to now there has been very few approaches for computer aided combination of results from different advanced techniques. This report presents the various mathematical fields involved towards that goal (statistical decision theory which allows the use of multiple hypothesis, non-linear decision theory for its capability to classify and to discriminate, graph theory to find the optimal path in an hypothesis graph and also fuzzy logic, multiple resolution analysis, artificial intelligence,...) and which combinations of methods are useful. Some images will illustrate this topic in which EDF is involved, and will explain what are the major goals of this work. Combining is not only an improvement of 3D visualisation which would allow to display simultaneously CAD or NDT data for example, but it consists in exploiting multisensor data collected via a variety of sophisticated techniques and presenting this information to the operator without overloading the operator/system capacities in order to reduce the uncertainty and to resolve the ambiguity inherent to mono method inspection. (author). 7 figs., 35 refs

  1. U.S. weapons-usable plutonium disposition policy: Implementation of the MOX fuel option

    Energy Technology Data Exchange (ETDEWEB)

    Woods, A.L. [ed.] [Amarillo National Resource Center for Plutonium, TX (United States); Gonzalez, V.L. [Texas A and M Univ., College Station, TX (United States). Dept. of Political Science

    1998-10-01

    A comprehensive case study was conducted on the policy problem of disposing of US weapons-grade plutonium, which has been declared surplus to strategic defense needs. Specifically, implementation of the mixed-oxide fuel disposition option was examined in the context of national and international nonproliferation policy, and in contrast to US plutonium policy. The study reveals numerous difficulties in achieving effective implementation of the mixed-oxide fuel option including unresolved licensing and regulatory issues, technological uncertainties, public opposition, potentially conflicting federal policies, and the need for international assurances of reciprocal plutonium disposition activities. It is believed that these difficulties can be resolved in time so that the implementation of the mixed-oxide fuel option can eventually be effective in accomplishing its policy objective.

  2. U.S. weapons-useable plutonium disposition policy: Implementation of the MOX fuel option

    International Nuclear Information System (INIS)

    Woods, A.L.; Gonzalez, V.L.

    1998-10-01

    A comprehensive case study was conducted on the policy problem of disposing of US weapons-grade plutonium, which has been declared surplus to strategic defense needs. Specifically, implementation of the mixed-oxide fuel disposition option was examined in the context of national and international nonproliferation policy, and in contrast to US plutonium policy. The study reveals numerous difficulties in achieving effective implementation of the mixed-oxide fuel option including unresolved licensing and regulatory issues, technological uncertainties, public opposition, potentially conflicting federal policies, and the need for international assurances of reciprocal plutonium disposition activities. It is believed that these difficulties can be resolved in time so that the implementation of the mixed-oxide fuel option can eventually be effective in accomplishing its policy objective

  3. Post-precipitations from MOX fuel solutions and analysis of microparticle formation in the PUREX process

    International Nuclear Information System (INIS)

    Henkelmann, R.; Baumgaertner, F.; Klein, F.; Niestroj, B.

    1989-01-01

    Subsequent precipitates of feed solutions from reprocessing were examined with the aid of the SEM-EDX method. On the one hand the examinations give information about the particle form and size distribution, on the other hand about the element distribution in single particles with consideration of the radiation data of the fuel. The subsequent precipitation samples which are examined in this study were taken after different residence times of the clarified fuel solutions. The examinations give information about the kind, element frequency, distribution and stoichiometry of single particles of the submicro- and microrange. (RB) [de

  4. Measurement of burnup in FBR MOX fuel irradiated to high burnup

    International Nuclear Information System (INIS)

    Koyama, Shin-ichi; Osaka, Masahiko; Sekine, Takashi; Morozumi, Katsufumi; Namekawa, Takashi; Itoh, Masahiko

    2003-01-01

    The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28% FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector. (author)

  5. A technique of melting temperature measurement and its application for irradiated high-burnup MOX fuels

    International Nuclear Information System (INIS)

    Namekawa, Takashi; Hirosawa, Takashi

    1999-01-01

    A melting temperature measurement technique for irradiated oxide fuels is described. In this technique, the melting temperature was determined from a thermal arrest on a heating curve of the specimen which was enclosed in a tungsten capsule to maintain constant chemical composition of the specimen during measurement. The measurement apparatus was installed in an alpha-tight steel box within a gamma-shielding cell and operated by remote handling. The temperature of the specimen was measured with a two-color pyrometer sighted on a black-body well at the bottom of the tungsten capsule. The diameter of the black-body well was optimized so that the uncertainties of measurement were reduced. To calibrate the measured temperature, two reference melting temperature materials, tantalum and molybdenum, were encapsulated and run before and after every oxide fuel test. The melting temperature data on fast reactor mixed oxide fuels irradiated up to 124 GWd/t were obtained. In addition, simulated high-burnup mixed oxide fuel up to 250 GWd/t by adding non-radioactive soluble fission products was examined. These data shows that the melting temperature decrease with increasing burnup and saturated at high burnup region. (author)

  6. Neutron coincidence counter for MOX fuel pins in storage trays: users' manual

    International Nuclear Information System (INIS)

    Cowder, L.; Menlove, H.

    1982-08-01

    The neutron coincidence counter for measurement of mixed-oxide fuel pins in storage trays is described. The special detector head has been designed so that the detectors, high-voltage junction boxes, and electronics are interchangeable with those of the high-level neutron coincidence counter system. This manual describes the system components and the operation and maintenance of the counter. The counter was developed at Los Alamos National Laboratory for in-plant inspection applications by the International Atomic Energy Agency

  7. Study of the lattice parameter evolution of PWR irradiated MOX fuel by X-Ray diffraction; Etude de l'evolution du parametre cristallin des combustibles MOX irradies en rep par la methode de diffraction des rayons X

    Energy Technology Data Exchange (ETDEWEB)

    Clavier, B

    1995-07-01

    Fuel irradiation leads to a swelling resulting from the formation of gaseous (Kr, Xe) or solid fission products which are found either in solution or as solid inclusions in the matrix. This phenomena has to be evaluated to be taken into account in fuel cladding Interaction. Fuel swelling was studied as a function of burn up by measuring the corresponding cell constant evolution by X-Ray diffraction. This study was realized on Mixed Oxide Fuels (MOX) irradiated in a Pressurized Water Reactor (PWR) at different burn-up for 3 initial Pu contents. Lattice parameter evolutions were followed as a function of burn-up for the irradiated fuel with and without an annealing thermal treatment. These experimental evolutions are compared to the theoretical evolutions calculated from the hard sphere model, using the fission product concentrations determined by the APPOLO computer code. Contribution of varying parameters influencing the unit cell value is discussed. Thermal treatment effects were checked by metallography, X-Ray diffraction and microprobe analysis. After thermal treatment, no structural change was observed but a decrease of the lattice parameter was measured. This modification results essentially from self-irradiation defect annealing and not from stoichiometry variations. Microprobe analysis showed that about 15% of the formed Molybdenum is in solid solution In the oxide matrix. Micrographs showed the existence of Pu packs in the oxide matrix which induces a broadening of diffraction lines. The RIETVELD method used to analyze the X-Ray patterns did not allow to characterize independently the Pu packs and the oxide matrix lattice parameters. Nevertheless, with this method, the presence of micro-strains in the irradiated nuclear fuel could be confirmed. (author)

  8. Development of a data base system for quality control of MOX fuels

    International Nuclear Information System (INIS)

    Takahashi, Kuniaki; Yamaguchi, Toshihiro; Mishima, Takeshi

    1988-01-01

    For the purpose of improvement and speed up of work concerning quality control for mixed oxide fuel fabrication, we have been developing a data base system having a data base consisting of data as to fabrication conditions and inspects, We have aimed to develop a data base system having capability of analysis and function giving informations as to quality control. The data base system is full interactive system on real time basis, consequently it makes analyzing and editing data easy. It has capability of relational research, numerical analysis, correlation analysis, drawing control charts, histograms, and other figures, and expressing status of fabrication processes using control charts. (author)

  9. Criticality evaluation of BWR MOX fuel transport packages using average Pu content

    International Nuclear Information System (INIS)

    Mattera, C.; Martinotti, B.

    2004-01-01

    Currently in France, criticality studies in transport configurations for Boiling Water Reactor Mixed Oxide fuel assemblies are based on conservative hypothesis assuming that all rods (Mixed Oxide (Uranium and Plutonium), Uranium Oxide, Uranium and Gadolinium Oxide rods) are Mixed Oxide rods with the same Plutonium-content, corresponding to the maximum value. In that way, the real heterogeneous mapping of the assembly is masked and covered by a homogeneous Plutonium-content assembly, enriched at the maximum value. As this calculation hypothesis is extremely conservative, COGEMA LOGISTICS has studied a new calculation method based on the average Plutonium-content in the criticality studies. The use of the average Plutonium-content instead of the real Plutonium-content profiles provides a highest reactivity value that makes it globally conservative. This method can be applied for all Boiling Water Reactor Mixed Oxide complete fuel assemblies of type 8 x 8, 9 x 9 and 10 x 10 which Plutonium-content in mass weight does not exceed 15%; it provides advantages which are discussed in our approach. With this new method, for the same package reactivity, the Pu-content allowed in the package design approval can be higher. The COGEMA LOGISTICS' new method allows, at the design stage, to optimise the basket, materials or geometry for higher payload, keeping the same reactivity

  10. Process control and safeguards system plutonium inventory conrol for MOX fuel facility

    International Nuclear Information System (INIS)

    Mishima, T.; Aoki, M.; Muto, T.; Amanuma, T.

    1979-01-01

    The plutonium inventory control (PINC) system is a real-time material accountability control system that is expected to be applied to a new large-scale plutonium fuel production facility for both fast breeder reactor and heavy water reactor at the Power Reactor and Nuclear Development Corporation. The PINC is basically a system for material control but is expected to develop into a whole facility control system, including criticality control, process control, quality control, facility protection, and so forth. Under PINC, every process and storage area is divided into a unit area, which is the smallest unit for both accountability and process control. Item and material weight automatically are accounted for at every unit area, and data are simultaneously treated by a computer network system. Sensors necessary for the system are being developed. 9 figures

  11. Modelling, simulation and visualisation for electromagnetic non-destructive testing

    International Nuclear Information System (INIS)

    Ilham Mukriz Zainal Abidin; Abdul Razak Hamzah

    2010-01-01

    This paper reviews the state-of-the art and the recent development of modelling, simulation and visualization for eddy current Non-Destructive Testing (NDT) technique. Simulation and visualization has aid in the design and development of electromagnetic sensors and imaging techniques and systems for Electromagnetic Non-Destructive Testing (ENDT); feature extraction and inverse problems for Quantitative Non-Destructive Testing (QNDT). After reviewing the state-of-the art of electromagnetic modelling and simulation, case studies of Research and Development in eddy current NDT technique via magnetic field mapping and thermography for eddy current distribution are discussed. (author)

  12. Material Accounting Issues at the U.S. MOX Fuel Fabrication Facility

    International Nuclear Information System (INIS)

    Lyman, E.

    2015-01-01

    The Mixed Oxide Fuel Fabrication Facility (MFFF) is under construction in the United States. The plant is being licenced by the U.S. Nuclear Regulatory Commission (NRC), which as the U.S. SSAC regulates both domestic MC and A and compliance with international safeguards (where applicable). Among the NRC's MC and A requirements for Category I fuel cycle facilities are programmes for item and process monitoring. The NRC also has requirements for timely resolution of alarms and assessment of the validity of alleged thefts. NRC's item monitoring requirement specifies that the operator must be able to verify the 'presence and integrity' of items, with the goal of detecting the loss of items containing 2 kilogrammes of plutonium within certain time periods. The requirements for resolution of alarms and assessment of alleged thefts also generally require some capability to locate and verify items on demand. However, to the extent these regulations mandate that individual items be physically located and verified by hand, they can be difficult (or impossible) to meet for facilities with large numbers of items. The MFFF design was based largely on French facilities that were not subject to similar requirements. Consequently, the applicant proposed a novel item monitoring approach that relies on the data within the plant's computerized inventory and process control systems. This proposal was challenged in July 2010 by intervenors, raising questions such as whether computer systems could be used as the sole means for verification, given the potential for data to be compromised. In February 2014, the NRC's Atomic Safety and Licencing Board issued a decision upholding the applicant's plan, but one of the three judges issued a dissent, citing concern about cyberterrorism. This paper will discuss the issues argued during the hearing and their broader relevance. (author)

  13. Joule-Heated Ceramic-Lined Melter to Vitrify Liquid Radioactive Wastes Containing Am241 Generated From MOX Fuel Fabrication in Russia

    International Nuclear Information System (INIS)

    Smith, E C; Bowan II, B W; Pegg, I; Jardine, L J

    2004-01-01

    The governments of the United Stated of America and the Russian Federation (RF) signed an Agreement September 1, 2000 to dispose of weapons plutonium that has been designated as no longer required for defense purposes. The Agreement declares that each country will disposition 34MT of excess weapons grade plutonium from their stockpiles. The preferred disposition technology is the fabrication of mixed oxide (MOx) fuel for use or burning in pressurized water reactors to destroy the plutonium. Implementation of this Agreement will require the conversion of plutonium metal to oxide and the fabrication of MOx fuel within the Russian Federation. The MOx fuel fabrication and metal to oxide conversion processes will generate solid and liquid radioactive wastes containing trace amounts of plutonium, neptunium, americium, and uranium requiring treatment, storage, and disposal. Unique to the Russian MOx fuel fabrication facility's flow-sheet is a liquid waste stream with high concentrations (∼1 g/l) of 241 Am and non radioactive silver. The silver is used to dissolve PuO 2 feed materials to the MOx fabrication facility. Technical solutions are needed to treat and solidify this liquid waste stream. Alternative treatment technologies for this liquid waste stream are being evaluated by a Russian engineering team. The technologies being evaluated include borosilicate and phosphate vitrification alternatives. The evaluations are being performed at a conceptual design level of detail under a Lawrence Livermore National Laboratory (LLNL) contract with the Russian organization TVEL using DOE NA-26 funding. As part of this contract, the RF team is evaluating the technical and economic feasibility of the US borosilicate glass vitrification technology based on a Duratek melter to solidify this waste stream into a form acceptable for storage and geologic disposal. The composition of the glass formed from treating the waste is dictated by the concentration of silver and americium it

  14. Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vapirev, E I; Jordanov, T; Christoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1994-12-31

    The idea of conversion of highly enriched uranium (HEU) from warheads without mixing it with natural uranium as well as the utilization of plutonium as fuel component is discussed. A nuclear fuel which is a mixture of 4% {sup 235}U (HEU) as a fissile isotope and 96 % {sup 232}Th (ThO{sub 2}) as a non-fissile isotope in a mixed oxide with thorium fuel is proposed. It is assumed that plutonium can also be used in the proposed fuel in a mixture with {sup 235}U. The following advantages of the use of HEU in LWRs in mixed {sup 235}U - Th fuel are pointed out: (1) No generation of long-living plutonium and americium isotopes (in case of reprocessing the high level radioactive wastes will contain only fission fragments and uranium); (2) The high conversion ratio of Th extends the expected burnup by approximately 1/3 without higher initial enrichment (the same initial enrichment simplifies the problem for compensation of the excess reactivity in the beginning with burnable poison and boric acid); (3) The high conversion ratio of Th allows the fuel utilization with less initial enrichment (by approx. 1/3) for the same burnup; thus less excess reactivity has to be compensated after reloading; in case of fuel reprocessing all fissile materials ({sup 235}U + {sup 233}U) could be chemically extracted. Irrespectively to the optimistic expectations outlined, further work including data on optimal loading and reloading schemes, theoretical calculations of thermal properties of {sup 235}U + Th fuel rods, manufacturing of several test fuel assemblies and investigations of their operational behaviour in a reactor core is still needed. 1 fig., 7 refs.

  15. Non-Destructive Radiological Characterisation Of Contaminated And Activated Concrete

    International Nuclear Information System (INIS)

    Cantrel, E.

    2005-01-01

    The decommissioning of nuclear facilities leads to large quantities of concrete and building material wastes. Radioactivity in building structures arise from very different physical processes such as neutron activation (bioshield), diffusion of the contaminants in the material (primary coolant leakage, maintenance and fuel loading) or aerosol deposition. The variety of the building material also extends the range of faced radiological characteristics. Therefore the minimization of concrete waste generation requires extensive characterisations and the availability of different measurement tools and methods. Up to now, these measurements came from the coring and the radiological analysis of the concrete, which is a destructive, long and costly technique. Looking for alternative solutions, SCK-CEN has started to investigate in collaboration with EDF -research and development and CEA (France) several non destructive methods based on gamma spectroscopy and different spectral examinations using mathematical calibration and modelling tools available on the market. Information on in-depth activity distribution can be derived from in situ gamma spectra by modelling absorption laws (peak to peak ratios) and photons interactions (Compton front) in the bulk of the concrete. As they combine modelling and measurement, the different methodologies being evaluated involve a lot of uncertainty sources linked to the measurement environment, to the knowledge available on site (historical background, material composition), to the operator responsible for the data acquisition and to the performance of the equipment. Therefore a detailed sensitivity analysis is required to define the range of applicability and the performances of the different methods

  16. Advanced non-destructive methods for an efficient service performance

    International Nuclear Information System (INIS)

    Rauschenbach, H.; Clossen-von Lanken Schulz, M.; Oberlin, R.

    2015-01-01

    Due to the power generation industry's desire to decrease outage time and extend inspection intervals for highly stressed turbine parts, advanced and reliable Non-destructive methods were developed by Siemens Non-destructive laboratory. Effective outage performance requires the optimized planning of all outage activities as well as modern Non-destructive examination methods, in order to examine the highly stressed components (turbine rotor, casings, valves, generator rotor) reliably and in short periods of access. This paper describes the experience of Siemens Energy with an ultrasonic Phased Array inspection technique for the inspection of radial entry pinned turbine blade roots. The developed inspection technique allows the ultrasonic inspection of steam turbine blades without blade removal. Furthermore advanced Non-destructive examination methods for joint bolts will be described, which offer a significant reduction of outage duration in comparison to conventional inspection techniques. (authors)

  17. Enhancement of actinide incineration and transmutation rates in Ads EAP-80 reactor core with MOX PuO2 and UO2 fuel

    International Nuclear Information System (INIS)

    Kaltcheva-Kouzminava, S.; Kuzminov, V.; Vecchi, M.

    2001-01-01

    Neutronics calculations of the accelerator driven reactor core EAP-80 with UO 2 and PuO 2 MOX fuel elements and Pb-Bi coolant are presented in this paper. Monte Carlo optimisation computations of several schemes of the EAP-80 core with different types of fuel assemblies containing burnable absorber B4 C or H 2 Zr zirconium hydride moderator were performed with the purpose to enhance the plutonium and actinide incineration rate. In the first scheme the reactor core contains burnable absorber B4 C arranged in the cladding of fuel elements with high enrichment of plutonium (up to 45%). In the second scheme H2 Zr zirconium hydride moderated zones were located in fuel elements with low enrichment (∼20%). In both schemes the incineration rate of plutonium is about two times higher than in the reference EAP-80 core and at the same time the power density distribution remains significantly unchanged compared to the reference core. A hybrid core containing two fuel zones one of which is the inner fuel region with UO 2 and PuO 2 high enrichment plutonium fuel and the second one is the outer region with fuel elements containing zirconium hydride layer was also considered. Evolution of neutronics parameters and actinide transmutation rates during the fuel burn-up is presented. Calculations were performed using the MCNP-4B code and the SCALE 4.3 computational system. (author)

  18. Non-destructive assay of radioactive waste

    International Nuclear Information System (INIS)

    Eid, C.; Bernard, P.

    1990-01-01

    The nuclear fuel cycle generates a large variety of waste containing Pu. After treatment and conditioning the final destination of this waste is either to be disposed by shallow land burial or in underground geological repositories. The method of disposal is determined by the quantity of Pu contained in the waste to be disposed of. For this reason and taking into account the rigorous requirements of the safety authorities concerning the protection of people and the environment, it is most important to determine accurately the Pu contents in the waste. Separate abstracts were prepared for 28 papers in this book

  19. Full MOX core for PWRs

    International Nuclear Information System (INIS)

    Puill, A.; Aniel-Buchheit, S.

    1997-01-01

    Plutonium management is a major problem of the back end of the fuel cycle. Fabrication costs must be reduced and plant operation simplified. The design of a full MOX PWR core would enable the number of reactors devoted to plutonium recycling to be reduced and fuel zoning to be eliminated. This paper is a contribution to the feasibility studies for achieving such a core without fundamental modification of the current design. In view of the differences observed between uranium and plutonium characteristics it seems necessary to reconsider the safety of a MOX-fuelled PWR. Reduction of the control worth and modification of the moderator density coefficient are the main consequences of using MOX fuel in a PWR. The core reactivity change during a draining or a cooling is thus of prime interest. The study of core global draining leads to the following conclusion: only plutonium fuels of very poor quality (i.e. with low fissile content) cannot be used in a 900 MWe PWR because of a positive global voiding reactivity effect. During a cooling accident, like an spurious opening of a secondary-side valve, the hypothetical return to criticality of a 100% MOX core controlled by means of 57 control rod clusters (made of hafnium-clad B 4 C rods with a 90% 10 B content) depends on the isotopic plutonium composition. But safety criteria can be complied with for all isotopic compositions provided the 10 B content of the soluble boron is increased to a value of 40%. Core global draining and cooling accidents do not present any major obstacle to the feasibility of a 100% MOX PWR, only minor hardware modifications will be required. (author)

  20. Vver-1000 Mox core computational benchmark

    International Nuclear Information System (INIS)

    2006-01-01

    The NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, fuel performance and fuel cycle issues related to disposing of weapons-grade plutonium in mixed-oxide fuel. The objectives of the group are to provide NEA member countries with up-to-date information on, and to develop consensus regarding, core and fuel cycle issues associated with burning weapons-grade plutonium in thermal water reactors (PWR, BWR, VVER-1000, CANDU) and fast reactors (BN-600). These issues concern core physics, fuel performance and reliability, and the capability and flexibility of thermal water reactors and fast reactors to dispose of weapons-grade plutonium in standard fuel cycles. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close co-operation (jointly, in most cases) with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A prominent part of these activities include benchmark studies. At the time of preparation of this report, the following benchmarks were completed or in progress: VENUS-2 MOX Core Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); VVER-1000 LEU and MOX Benchmark (completed); KRITZ-2 Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); Hollow and Solid MOX Fuel Behaviour Benchmark (completed); PRIMO MOX Fuel Performance Benchmark (ongoing); VENUS-2 MOX-fuelled Reactor Dosimetry Calculation (ongoing); VVER-1000 In-core Self-powered Neutron Detector Calculational Benchmark (started); MOX Fuel Rod Behaviour in Fast Power Pulse Conditions (started); Benchmark on the VENUS Plutonium Recycling Experiments Configuration 7 (started). This report describes the detailed results of the benchmark investigating the physics of a whole VVER-1000 reactor core using two-thirds low-enriched uranium (LEU) and one-third MOX fuel. It contributes to the computer code certification process and to the

  1. Contributions to ultrasounds applications in non-destructive tests on materials used in nuclear technologies

    International Nuclear Information System (INIS)

    Stanica, V.

    1979-01-01

    The problems expounded in the paper, besides servjng the practical purpose generated by the need to perform quality tests on fuel element compounds by means of the ultrasounds method, are also interesting to ultrasounds non-destructive tests applied in all industry branches as they assert the necessity of passing from manual to automation tests carried out by installations which should record the signals caused by failures, both to increase the productivjty and especially to transform it into an objective, effective test. (author)

  2. MOX - equilibrium core design and trial irradiation in KAPS - 1

    International Nuclear Information System (INIS)

    Pradhan, A.S.; Ray, Sherly; Kumar, A.N.; Parikh, M.V.

    2006-01-01

    Option of usage of MOX fuel bundles in the equilibrium core of Indian 220 MWe PHWRs on a regular basis has been studied. The design of the MOX bundle considered is MOX -7 with inner 7 elements with uranium and plutonium oxide MOX fuel and outer 12 elements with natural uranium fuel. The composition of the plutonium isotopes corresponds to that at about 6500 MWD/TeU burnup. Burnup optimization has been done such that operation at design rated power is possible while achieving the maximum average discharge burnup. Operation with the optimized burnup pattern will result in substantial saving of natural uranium bundles. To obtain feedback on the performance of MOX bundles prior to its large scale use about 50 MOX-7 bundles have been loaded in KAPS - 1 equilibrium core. Locations have been selected such that reactor should be operating at rated power without violating any constraints on channel bundle powers and also meeting the safety requirements. Burnup of interest also should be achieved in minimum period of time. The fissile plutonium content in the 50 MOX fuel bundles loaded is about 75.6 wt % . About 38 bundles out of the 50 bundles loaded have been already discharged and remaining bundles are still in the core. The maximum discharge burnup of the MOX bundles is about 12000 MWD/TeU. The performance of the MOX bundles were excellent and as per prediction. No MOX bundle is reported to be failed. (author)

  3. The status of BNFL's MOX project

    International Nuclear Information System (INIS)

    Edwars, John; Cooch, Julian P.; Slater, Michel W.

    2002-01-01

    Full text: In the late 1980s BNFL decided to enter the MOX fuel fabrication business to support our reprocessing business and return the plutonium product to our customers in the useable form of MOX fuel. The first phase of the strategy was to gain some irradiation experience for MOX produced by our own Short Binderless Route (SBR) process. To achieve this the MOX Demonstration Facility (MDF) was built at Sellafield and 28 MOX fuel assemblies were produced up to 1998 that were loaded into PWRs in Europe. In 1994, BNFL started the construction of their large scale MOX production plant, SMP. The design and construction of the plant and supporting facilities was completed some years ago and the commissioning of the plant with uranium commenced around June 1999. In October 2001, the UK Government provided BNFL with the approval to operate SMP with plutonium. On 20 December 2001, the UK Regulators gave BNFL their approval to start plutonium operations. This paper summarises the approach used to commission SMP and describes some of the lessons learnt during the commissioning phase of the project and the start up of the plant with plutonium. An explanation of our experience obtaining a licence to operate the plant is provided together with a description of the changes we have made to ensure that the quality of the product from SMP can be guaranteed. Finally, the paper summarises the experience BNFL has gained during irradiating MOX fuel produced by the SBR process and explains how the data compares with that available for UO2 and supports the in reactor use of MOX fuel made in SMP. (author)

  4. AP1000 core design with 50% MOX loading

    International Nuclear Information System (INIS)

    Fetterman, Robert J.

    2009-01-01

    The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core design and a mixed MOX/UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance

  5. AP1000 core design with 50% MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)

    2008-07-01

    The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core and a mixed MOX / UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

  6. AP1000 core design with 50% MOX loading

    International Nuclear Information System (INIS)

    Fetterman, Robert J.

    2008-01-01

    The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core and a mixed MOX / UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

  7. AP1000 core design with 50% MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)], E-mail: fetterrj@westinghouse.com

    2009-04-15

    The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core design and a mixed MOX/UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance.

  8. Non-destructive characterization of the materials for future nuclear reactors

    International Nuclear Information System (INIS)

    Snopek, J.; Slugen, V.

    2016-01-01

    For our experiments, we have used Barkhausen noise technique, which is powerful non-destructive method for monitoring stresses in lattices of magnetic materials. We have also used PAS, which is powerful non-destructive method for diagnosing vacancy defects in variable materials. We researched some ODS steels, which are primarily going to be used as fuel cladding or reactor pressure vessel internal components. This thesis describes the behavior of the microstructure of the oxide dispersion strengthened steels at intermediate temperature. Two, in principle, different techniques were used for the characterization of the microstructure of the oxide dispersion strengthened steels thermally aged at 475 grad C and 650 grad C. Both techniques, namely Positron annihilation lifetime spectroscopy (PAS) and Barkhausen noise (BN) measurements are very sensitive to metallurgical modifications and presence of nano-sized imperfections in the crystal lattice. Precipitation of the nano-sized α' phases shift the Barkhausen noise signal. (authors)

  9. Non-destructive testing and evaluation for structural integrity

    International Nuclear Information System (INIS)

    Baldev Raj; Jayakumar, T.; Rao, B.P.C.

    1995-01-01

    In this paper, a brief description of the physical concepts of non-destructive evaluation (NDE) methods and the physical/derived parameters that are used for assessing defects, stresses and microstructures are given. A few case studies highlighting the importance of non-destructive testing and evaluation for structural integrity assessment are also discussed based on the investigations carried out. Emerging concepts like intelligent processing of materials, expert systems, neural networks, use of multisensors with fusion of data and exploitation of signal analysis and imaging approaches are also addressed in this paper. (author). 92 refs., 1 tab

  10. Magnetic Non-destructive Testing of Plastically Deformed Mild Steel

    Directory of Open Access Journals (Sweden)

    Jozef Pala

    2004-01-01

    Full Text Available The Barkhausen noise analysis and coercive field measurement have been used as magnetic non-destructive testing methods for plastically deformed high quality carbon steel specimens. The strain dependence of root mean square value and power spectrum of the Barkhausen noise and the coercive field are explained in terms of the dislocation density. The specimens have been subjected to different magnetizing frequencies to show the overlapping nature of the Barkhausen noise. The results are discussed in the context of usage of magnetic non-destructive testing to evaluate the plastic deformation of high quality carbon steel products.

  11. Use of fission track analysis technique for the determination of MicroBequerel level of {sup 239}Pu in urine samples from radiation workers handling MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yadav, J.R., E-mail: yadav_jogendra@rediffmail.co [Health Physics Laboratory, Health Physics Division, BARC, Tarapur 401502 (India); Rao, D.D.; Kumar, Ranjeet [Health Physics Laboratory, Health Physics Division, BARC, Tarapur 401502 (India); Aggarwal, S.K. [Fuel chemistry Division, BARC, Trombay, Mumbai 400085 (India)

    2011-07-15

    Fission track analysis (FTA) technique for the determination of {sup 239}Pu excreted through urine has been standardized using blank samples, tracer and {sup 239}Pu spikes. Double stage anion exchange separation protocol has been applied and an average radiochemical recovery of {sup 239}Pu of 18% was obtained. An average track registration efficiency of 11 tracks per {mu}Bq of {sup 239}Pu, irradiated to 0.35x10{sup 17} neutron fluence was established. Reagent blank urine samples from 11 controlled subjects were analyzed by FTA and an average of 149{+-}14 tracks was obtained. Minimum detectable activity of 34 {mu}Bq L{sup -1} of urine sample was obtained and will be useful for monitoring chronic exposure cases handling MOX fuel.

  12. Modelling and simulation of eddy current non-destructive testing

    International Nuclear Information System (INIS)

    Mansir, H.; Burais, N.; Nicolas, A.

    1986-01-01

    This paper presents the practical configuration for detecting cracks in conducting materials by eddy current non destructive testing. An electromagnetic field formulation is proposed using Maxwell's relations. Geometrical and physical properties of the crack are taken into account by several models, particularly with a new finite element called ''crack element''. Modelisation is applied to sensor impedance calculation with classical numerical methods [fr

  13. Non-destructive testing of tubes by electromagnetic processes

    International Nuclear Information System (INIS)

    Kowarski, A.

    1979-01-01

    This article reviews and assesses the non destructive testing techniques used for locating defects in tubes by electromagnetic processes. These form the basis of many testing devices, the diversity of which results from various factors: range of materials, methods of fabrication, specific defects of the product. There are two distinct main families of devices utilising two different principles: dispersion flow and Foucault currents [fr

  14. Mathematical modelling of ultrasonic non-destructive evaluation

    Directory of Open Access Journals (Sweden)

    Larissa Ju Fradkin

    2001-01-01

    Full Text Available High-frequency asymptotics have been used at our Centre to develop codes for modelling pulse propagation and scattering in the near-field of the ultrasonic transducers used in NDE (Non-Destructive Evaluation, particularly of walls of nuclear reactors. The codes are hundreds of times faster than the direct numerical codes but no less accurate.

  15. Online Preventative Non-Destructive Evaluation in Automated Fibre Placement

    NARCIS (Netherlands)

    Tonnaer, R.; Shroff, S.; Groves, R.M.

    2016-01-01

    The strict quality requirements for aerospace composite struc- tures give rise to costly quality control procedures. In automated bre placement (AFP) these procedures rely heavily on manual work and inspection. This research aims at performing preventative non-destructive evaluation of composite

  16. Non-destructive controls in the steel tube industry

    International Nuclear Information System (INIS)

    Mondot, J.

    1978-01-01

    The main non-destructive control methods in the tube industry are reviewed: eddy currents, particularly well adapted to small tubes; magnetoscopic testing for weldless tubes; ultrasonic waves widely used for thick weldless tubes and weldings; radiography, to examine tube ends and the known questionable zones; measure of diameters by laser [fr

  17. NonDestructive Evaluation for Industrial & Development Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hunter, James F. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-10-12

    Provide overview of weld inspection for Non-Destructive Testing at LANL. This includes radiography (RT/DR/CR/CT for x-ray & neutron sources), ultrasonic testing (UT/PAUT), dye penetrant inspection (PT), eddy current inspection (ET) and magnetic particle testing (MT). Facilities and capabilities for weld inspection will be summarized with examples.

  18. Non-destructive controls in the mechanical industry

    Energy Technology Data Exchange (ETDEWEB)

    Jarlan, L

    1978-12-01

    The sequence of operations implicating the mechanical industries from the suppliers to their customers is briefly recalled; a description of the field of application of non-destructive control methods in these industries is given. Follows a description of some recent typical applications of the principal methods: radiography, ultrasonic waves, magnetism, acoustic emission, sonic control, tracer techniques.

  19. Laser plasma X-ray for non-destructive inspection

    International Nuclear Information System (INIS)

    Yagi, T.; Kusama, H.

    1995-01-01

    External electric field is applied to the laser produced plasma, and its found that plasma shape in soft X-ray region is changed due to the penetrating electric field. The plasma emits strong hard X-ray, which can be used as a compact light source for non-destructive inspection. (author)

  20. Non-destructive Faraday imaging of dynamically controlled ultracold atoms

    DEFF Research Database (Denmark)

    Gajdacz, Miroslav; Pedersen, Poul Lindholm; Mørch, Troels

    2013-01-01

    We describe an easily implementable method for non-destructive measurements of ultracold atomic clouds based on dark field imaging of spatially resolved Faraday rotation. The signal-to-noise ratio is analyzed theoretically and, in the absence of experimental imperfections, the sensitivity limit...

  1. Non-destructive testing of electronic component packages

    International Nuclear Information System (INIS)

    Anderle, C.

    1975-01-01

    A non-destructive method of investigating packaged parts of semiconductor components by X radiation is described and the relevant theoretical relations limiting this technique are derived. The application of the technique is demonstrated in testing several components. The described method is iNsimple and quick. (author)

  2. Training Guidelines in Non-destructive Testing Techniques. 2013 Edition

    International Nuclear Information System (INIS)

    2014-12-01

    The IAEA promotes industrial applications of radiation technology, including non-destructive testing (NDT), through activities such as Technical Cooperation Projects (national and regional) and Coordinated Research Projects. Through this cooperation, Member States have initiated national programmes for the training and certification of NDT personnel. National certifying bodies have also been established based on International Organization for Standardization (ISO) standards. As part of these efforts, the IAEA has been actively involved in developing training materials. Consequently, IAEA-TECDOC-407, Training Guidelines in Non-destructive Testing Techniques, was published in 1987, then revised and expanded as IAEA-TECDOC-628 in 1991. Revisions of IAEA-TECDOC-628 were considered essential to meet the demands of end-user industries in Member States, and revised and expanded versions were issued in 2002 and 2008. These latter versions included work conducted by the International Committee for Non-Destructive Testing (ICNDT) and many national NDT societies. It is one of the publications referred to in ISO 9712:2005, Non-destructive Testing: Qualification and Certification of Personnel, which in turn is an internationally accepted standard, revised as ISO 9712:2012, Non-destructive Testing: Qualification and Certification of NDT Personnel. This publication is an updated version of IAEA-TECDOC-628. The content of which has been revised following the changes of ISO 9712 converging with EN 473 and becoming EN ISO 9712:2012, based on the experience of experts and comments from end-user industries. The details of the topics on each subject have been expanded to include the latest developments in the respective methods. The incorporated changes will assist the end-user industries to update their NDT qualification and certification schemes and course materials. This publication, like the previous versions, will continue to play an important role in international harmonization

  3. In situ non-destructive measurement of biofilm thickness and topology in an interferometric optical microscope

    Energy Technology Data Exchange (ETDEWEB)

    Larimer, Curtis [Pacific Northwest National Laboratory, Battelle for the USDOE, PO Box 999, MSIN P7-50 Richland WA 99354 USA; Suter, Jonathan D. [Pacific Northwest National Laboratory, Battelle for the USDOE, PO Box 999, MSIN P7-50 Richland WA 99354 USA; Bonheyo, George [Pacific Northwest National Laboratory, Battelle for the USDOE, PO Box 999, MSIN P7-50 Richland WA 99354 USA; Addleman, Raymond Shane [Pacific Northwest National Laboratory, Battelle for the USDOE, PO Box 999, MSIN P7-50 Richland WA 99354 USA

    2016-03-15

    Biofilms are ubiquitous and deleteriously impact a wide range of industrial processes, medical and dental health issues, and environmental problems such as transport of invasive species and the fuel efficiency of ocean going vessels. Biofilms are difficult to characterize when fully hydrated, especially in a non-destructive manner, because of their soft structure and water-like bulk properties. Herein we describe a non-destructive high resolution method of measuring and monitoring the thickness and topology of live biofilms of using white light interferometric optical microscopy. Using this technique, surface morphology, surface roughness, and biofilm thickness can be measured non-destructively and with high resolution as a function of time without disruption of the biofilm activity and processes. The thickness and surface topology of a P. putida biofilm were monitored growing from initial colonization to a mature biofilm. Typical bacterial growth curves were observed. Increase in surface roughness was a leading indicator of biofilm growth.

  4. MOX recycling-an industrial reality

    International Nuclear Information System (INIS)

    Shallo, G.D.F.

    1996-01-01

    Reprocessing and plutonium recycling have now attained industrial maturity in France and Europe. Specifically, mixed-oxide (MOX) fuel is fabricated and used in light water reactors (LWRs) in satisfactory operating conditions. The utilities and the fuel cycle industry experience no technical difficulties, and European recycling programs are growing steadily, from 18 reactors in operation today up to 50 expected around the year 2000, putting the system reprocessing-recycling in coherence: 25 t of plutonium will then be used each year to produce the electricity equivalence of 25 millions tons of oil. Plutonium recycling in MOX fuel in current LWRs proves to be technically safe and economically competitive and meets natural resource savings and environmental protection objectives. And recycling responds properly to the nonproliferation concerns. Such an industrial experience gives a unique reference for weapons plutonium disposition through MOX use in reactors

  5. The MOX Demonstration Facility - the stepping stone to commercial MOX production

    International Nuclear Information System (INIS)

    Macdonald, A.G.

    1994-01-01

    The paper provides an insight into MOX fuel and the economic benefits of its use in pressurized water reactors (PWRs). BNFL and AEA are collaborating in the design, construction and operation of a thermal MOX Demonstration Facility (MDF) on the AEA Windscale site in Cumbria. The process flowsheet and equipment employed in MDF are discussed and the special precautions required to handle plutonium bearing materials are highlighted. The process flowsheet includes the short binderless route which has been specially developed for use in MDF and results in fuel pellets with an homogeneous structure. MDF is the forerunner to the design and construction of a larger scale Sellafield MOX Plant and hence is the stepping-stone to commercial MOX production. (author)

  6. Non-destructive microstructural analysis with depth resolution

    Energy Technology Data Exchange (ETDEWEB)

    Zolotoyabko, E. E-mail: zloto@tx.technion.ac.il; Quintana, J.P

    2003-01-01

    A depth-sensitive X-ray diffraction technique has been developed with the aim of studying microstructural modifications in inhomogeneous polycrystalline materials. In that method, diffraction profiles are measured at different X-ray energies varied by small steps. X-rays at higher energies probe deeper layers of material. Depth-resolved structural information is retrieved by comparing energy-dependent diffraction profiles. The method provides non-destructive depth profiling of the preferred orientation, grain size, microstrain fluctuations and residual strains. This technique is applied to the characterization of seashells. Similarly, energy-variable X-ray diffraction can be used for the non-destructive characterization of different laminated structures and composite materials.

  7. Non-Destructive Techniques Based on Eddy Current Testing

    Science.gov (United States)

    García-Martín, Javier; Gómez-Gil, Jaime; Vázquez-Sánchez, Ernesto

    2011-01-01

    Non-destructive techniques are used widely in the metal industry in order to control the quality of materials. Eddy current testing is one of the most extensively used non-destructive techniques for inspecting electrically conductive materials at very high speeds that does not require any contact between the test piece and the sensor. This paper includes an overview of the fundamentals and main variables of eddy current testing. It also describes the state-of-the-art sensors and modern techniques such as multi-frequency and pulsed systems. Recent advances in complex models towards solving crack-sensor interaction, developments in instrumentation due to advances in electronic devices, and the evolution of data processing suggest that eddy current testing systems will be increasingly used in the future. PMID:22163754

  8. Non-destructive elecrochemical monitoring of reinforcement corrosion

    DEFF Research Database (Denmark)

    Nygaard, Peter Vagn

    been widely accepted as a non-destructive ”state of the art” technique for detection of corrosion in concrete structures. And, over the last decade, the trend in corrosion monitoring has moved towards quantitative non-destructive monitoring of the corrosion rate of the steel reinforcement. A few...... corrosion rate measurement instruments have been developed and are commercially available. The main features of these instruments are the combined use of an electrochemical technique for determining the corrosion rate and a so-called ”confinement technique”, which in principle controls the polarised surface...... area of the reinforcement, i.e. the measurement area. Both on-site investigations and laboratory studies have shown that varying corrosion rates are obtained when the various commercially available instruments are used. And in the published studies, conflicting explanations are given illustrating...

  9. Non-destructive examination system of vitreous body

    Science.gov (United States)

    Shibata, Takuma; Gong, Jin; Watanabe, Yosuke; Kabir, M. Hasnat; Masato, Makino; Furukawa, Hidemitsu; Nishitsuka, Koichi

    2014-04-01

    Eyeball plays a quite important role in acquiring the vision. Vitreous body occupies the largest part of the eyeball and consists of biological, elastic, transparent, gel materials. In the present medical examination, the non-destructive examination method of the vitreous body has not been well established. Here, we focus on an application of dynamic light scattering to this topic. We tried to apply our lab-made apparatus, scanning microscopic light scattering (SMILS), which was specially designed for observing the nanometer-scale network structure in gel materials. In order to examine the vitreous body using SMILS method, a commercial apparatus, nano Partica (Horiba Co. Ltd.) was also customized. We analyzed vitreous body using both the SMILS and the customized nano Partica. We successfully examined the vitreous bodies of healthy pigs in non-destructive way.

  10. Advanced uses of radiation in non-destructive evaluation

    International Nuclear Information System (INIS)

    Baldev Raj; Viswanathan, B.; Venkataraman, B.

    1998-01-01

    The increasing demand for newer materials and stringency of specifications, have expanded the scope of advanced uses of radiation in non-destructive evaluation of materials and industrial components. This paper highlights the application of some of the advanced techniques of radiography and residual stress measurements, using x-ray diffraction, for materials characterisation and testing, based on the results obtained at the author's laboratory. The application of positron annihilation techniques based on the use of radioisotopes and high resolution gamma ray spectroscopy, is introduced as non-destructive tools for materials characterisation. Selective examples of significant results obtained using this technique, on the radiation damage and early stages of fatigue damage in technologically important steels are reviewed from recent works at the author's laboratory and elsewhere. The scope of application of charge particle based thin layer activation method is briefly outlined. (author)

  11. Non destructive testing of green parts in powder metallurgy

    International Nuclear Information System (INIS)

    Accary, A.

    1979-01-01

    The non destructive testing of green parts is potentially advantageous by making possible a lowering of the material and energy consumption as well as the production of parts with a 100% reliability. After a survey of the possible methods and of the defects to be detected it is shown that the goal can be achieved using a 'blind detection' method and that the difficulty of the problem depends on the size and shape of the part to be controled. The gravimetric, dimensional, γ absorption and thermal diffusivity methods are then examined. It is concluded that a unit control is paying only if it allows to enter the high reliability part market. Used statisticaly the non destructive testing of green parts can possibly lead to savings on materials and energy [fr

  12. The non-destructive control, a major constituent of quality

    International Nuclear Information System (INIS)

    Anon.

    1998-01-01

    The number of continuous research and development works about non-destructive control in all sectors of activity is justified by the increasing need for high quality materials without anomalies. This paper gives a overview of the state of the art and of the recent trends in non-destructive testing researches in different sectors: aeronautics, nuclear industry, automotive industry. New studies and techniques are presented: ultrasonic testing of welds on large diameter pipes, automated applications of ultrasonic testing, ultrasound/computer-aided design coupling, pressure vessels inspection using acoustic emission testing (leaks detection, application to composite materials), numerical radiography (image visualisation and processing), magnetic testing (steel damage detection using Barkhausen noise testing), 'shearography' (detection of the loss of thickness in pipes due to corrosion), X-ray tomography (density measurement of sintered steels, fluid flow calculations in automobile parts). (J.S.)

  13. Plutonium - out of the stockpile and into the MOX market

    International Nuclear Information System (INIS)

    Edwards, J.; Hexter, B.C.; Powell, D.J.

    1993-01-01

    Reducing the risks associated with growing stocks of plutonium is just one of the factors behind the manufacture of mixed oxide (MOX) fuel. A United Kingdom collaboration, described here, has recently taken the first steps into the market place for MOX. (Author)

  14. A non-destructive DNA sampling technique for herbarium specimens.

    Science.gov (United States)

    Shepherd, Lara D

    2017-01-01

    Herbarium specimens are an important source of DNA for plant research but current sampling methods require the removal of material for DNA extraction. This is undesirable for irreplaceable specimens such as rare species or type material. Here I present the first non-destructive sampling method for extracting DNA from herbarium specimens. DNA was successfully retrieved from robust leaves and/or stems of herbarium specimens up to 73 years old.

  15. Non-destructive tests of capsules for JMTR irradiation examination

    International Nuclear Information System (INIS)

    Tanaka, Hidetaka; Nagao, Yoshiharu; Sato, Masashi; Osawa, Kenji

    2007-03-01

    Irradiation examination are increasing in advanced irradiation research for accurate prediction control and evaluation of irradiation parameter such as neutron fluence, etc. by using JMTR. Irradiation capsule internals are therefore structurally complicated recently. This report described the procedure of non destructive tests such as radiographic test, penetrant test, ultrasonic test, etc. for inspection of irradiation capsules in JMTR, and the result of Test-case of confirmation procedure for internal parts of irradiation capsules. (author)

  16. Non destructive testing and neutron radiography in 1994

    International Nuclear Information System (INIS)

    Bayon, G.

    1994-01-01

    Neutron radiography has been considered for a long time as a promising technique; however it plays a minor part in the world of non destructive testing today, due to the lack of suitable neutron sources and lack of new industrial applications. This paper reviews the present status of neutron sources, neutron radiography activities, especially in France (such as the neutron-capture-issued secondary radiation spectrometry), in nuclear, aerospace, aeronautical and metallurgical sectors, and the last applications of neutron dynamic imaging. 9 refs

  17. Guidebook on non-destructive testing of concrete structures

    International Nuclear Information System (INIS)

    2002-01-01

    The International Atomic Energy Agency has been active in the promotion of non-destructive testing (NDT) technology for many years. NDT is an important component of a number of IAEA regional projects. This guidebook deals with NDT of concrete. This book covers a wide range of NDT methods including industrial radiography, ultrasonic testing, electromagnetic testing, infrared thermography, etc. Codes, standards, specifications and procedures are also covered

  18. Non-Destructive Testing for Control of Radioactive Waste Package

    Science.gov (United States)

    Plumeri, S.; Carrel, F.

    2015-10-01

    Characterization and control of radioactive waste packages are important issues in the management of a radioactive waste repository. Therefore, Andra performs quality control inspection on radwaste package before disposal to ensure the compliance of the radwast characteristics with Andra waste disposal specifications and to check the consistency between Andra measurements results and producer declared properties. Objectives of this quality control are: assessment and improvement of producer radwaste packages quality mastery, guarantee of the radwaste disposal safety, maintain of the public confidence. To control radiological characteristics of radwaste package, non-destructive passive methods (gamma spectrometry and neutrons counting) are commonly used. These passive methods may not be sufficient, for instance to control the mass of fissile material contained inside radwaste package. This is particularly true for large concrete hull of heterogeneous radwaste containing several actinides mixed with fission products like 137Cs. Non-destructive active methods, like measurement of photofission delayed neutrons, allow to quantify the global mass of actinides and is a promising method to quantify mass of fissile material. Andra has performed different non-destructive measurements on concrete intermediate-level short lived nuclear waste (ILW-SL) package to control its nuclear material content. These tests have allowed Andra to have a first evaluation of the performance of photofission delayed neutron measurement and to identify development needed to have a reliable method, especially for fissile material mass control in intermediate-level long lived waste package.

  19. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    Energy Technology Data Exchange (ETDEWEB)

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A. [AREVA - Tour AREVA, 1 Place Jean Millier, 92084 Paris La Defense (France)

    2013-07-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

  20. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    International Nuclear Information System (INIS)

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

    2013-01-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO 2 fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory

  1. Summary of the Minor Actinide-bearing MOX AFC-2C and -2D Irradiations

    International Nuclear Information System (INIS)

    McClellan, Kenneth; Chichester, Heather; Hayes, Steve; Voit, Stewart

    2013-01-01

    Summary of AFC-2C and AFC-2D tests: • AFC-2C and 2D, 1st MOX experiments in FCRD, were irradiated in ATR; • Initial results indicate performance of experimental MA-MOX fuels are similar to standard FR MOX fuels; • Cd-shrouded ATR experiment assembly and 235 U enrichment produce prototypic fast reactor power and temperature profiles leading to classic MOX zone restructuring; • Baseline postirradiation examinations have been completed for AFC-2C MOX and MA-MOX fuels; • Future work includes: – PIE of AFC-2D; – compare results to prototypic MOX fuel performance; – electron microscopy for microstructure and constituent distribution; – advanced NDE on saved pins

  2. Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO2 and MOX spent fuels

    International Nuclear Information System (INIS)

    Sasahara, Akihiro; Matsumura, Tetsuo; Nicolaou, G.; Betti, M.; Walker, C.T.

    1997-01-01

    The post irradiation examinations (PIE) were carried out for high burn-up UO 2 spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, a) isotopic analysis, b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)

  3. Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO{sub 2} and MOX spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sasahara, Akihiro; Matsumura, Tetsuo [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.; Nicolaou, G.; Betti, M.; Walker, C.T.

    1997-03-01

    The post irradiation examinations (PIE) were carried out for high burn-up UO{sub 2} spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, (a) isotopic analysis, (b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)

  4. Design Studies of ''Island'' Type MOX Lead Test Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovitchev, A.M.

    2000-03-31

    In this document the results of neutronics studies of <> type MOX LTA design are presented. The characteristics both for infinite MOX grids and for VVER-1000 core with 3 MOX LTAs are calculated. the neutronics parameters of MOX fueled core have been performed using the Russian 3D code BIPR-7A and 2D code PERMAK-A with the constants prepared by the cell spectrum code TVS-M.

  5. Design of a reactor core in the Oma Full MOX-ABWR

    International Nuclear Information System (INIS)

    Hama, Teruo

    1999-01-01

    The Electric Power Development Co., Ltd. has progressed a construction plan on an improved boiling-water reactor aiming at loading of MOX fuel in all reactor cores (full MOX-ABWR) at Oma-cho, Aomori prefecture, which is a last stage on application of approval on establishment at present. Here were described on outlines of reactor core in the full MOX-ABWR and its safety evaluation. For the full MOX-ABWR loading MOX fuel assembly into all reactor core, thermal and mechanical design analysis of fuel bars and core design analysis were conducted. As a result, it was confirmed that judgement standards in mixed core of MOX fuel and uranium fuel were also applicable as well as that in uranium fuel. (G.K.)

  6. Non-destructive evaluation on mechanical properties of nuclear graphite with porous structure

    International Nuclear Information System (INIS)

    Shibata, Taiju; Hanawa, Satoshi; Sumita, Junya; Tada, Tatsuya; Sawa, Kazuhiro; Iyoku, Tatsuo

    2005-01-01

    As a research subjects of 'Research and development for advanced high temperature gas cooled reactor fuels and graphite components,' we started the study of development of non-destructive evaluation methods for mechanical properties of graphite components. The micro-indentation and ultrasonic wave methods are focused to evaluate the degradation of graphite components in VHTR core. For the micro-indentation method, the test apparatus was designed for the indentation test on graphite specimens with some stress levels. It is expected the stress condition is evaluated by the indentation load-depth characteristics and hardness. For the ultrasonic wave method, ultrasonic wave testing machine and probes were prepared for experiments. It is expected that the stress and inner porous conditions are evaluated by the wave propagation characteristics with wave-pore interaction model. R and D plan to develop the non-destructive evaluation method for graphite is presented in this paper. (This study is the result of contract research in the fiscal year of 2004, Research and development for advanced high temperature gas cooled reactor fuels and graphite components,' which is entrusted to the Japan Atomic Energy Research Institute from the Ministry of Education, Culture, Sports, Science and Technology of Japan.) (author)

  7. Isotope techniques in non-destructive testing of dynamic systems

    International Nuclear Information System (INIS)

    Singh, Gursharan; Pant, H.J.

    1996-01-01

    A few applications of gamma scanning and radiotracer techniques for Non-destructive Testing (NDT) of dynamic systems in chemical and petrochemical industries are briefly discussed in this paper. Examples of gamma scanning inspections carried out for troubleshooting of various types of columns such as vacuum, extraction, separator and rectifier, with trays and packed beds and having diameters from 1 meter to 8.4 meters are given. Radiotracer applications for Residence Time Distribution (RTD) studies on different systems like an aniline production reactor in a chemical industry and a laboratory scale solid-liquid fluidized bed column are mentioned. (author)

  8. Non-destructive characterization using pulsed fast-thermal neutrons

    International Nuclear Information System (INIS)

    Womble, P.C.

    1995-01-01

    Explosives, illicit drugs, and other contraband materials contain various chemical elements in quantities and ratios that differentiate them from each other and from innocuous substances. Furthermore, the major chemical elements in coal can provide information about various parameters of importance to the coal industry. In both examples, the non-destructive identification of chemical elements can be performed using pulsed fast-thermal neutrons that, through nuclear reactions, excite the nuclei of the various elements. This technique is being currently developed for the dismantling of nuclear weapons classified as trainers, and for the on-line coal bulk analysis. (orig.)

  9. Non-destructive testing and radiation in industry

    International Nuclear Information System (INIS)

    Woodford, C; Ashby, P.

    2001-01-01

    Non-destructive testing (NDT) is a little known discipline which uses non-invasive and passive techniques to investigate the condition of materials and structures. Some of these techniques employ the use of radioisotopes. The penetrating radiations produced by these materials are applied in various ways to obtain the required information. This presentation is an overview of the application of radioisotopes within the scope of NDT. Notwithstanding the well established use of traditional materials, new forms of radioisotopes are being developed which will extend their capabilities

  10. APPARATUS FOR NON-DESTRUCTIVE INSPECTION OF CANTILEVERED MEMBERS

    Science.gov (United States)

    Taylor, E.R.; Mahoney, C.H.; Lay, C.R.

    1961-10-24

    An apparatus for non-destructive inspection of cantilevered members, such as compressor blades, is described. The member under inspection is vibrated with a regulated source of air under pressure. The amplitude of vibration of the member is maintained at its natural frequency. The frequency of vibration of the member is measured. An indication of an excessive decay or erratic shifting in the measured frequency above an allowable hysteretic decay is provided as an indication of a fault in the member. The member is vibrated for a selected test period. (AEC)

  11. Non-destructive beam profile monitor at HIMAC

    International Nuclear Information System (INIS)

    Sato, S.; Araki, N.; Hosaka, M.

    1995-01-01

    Non-destructive profile monitors (NDPM), based on micro-channel plate (MCP), have been developed and installed in both the synchrotron ring and high-energy beam transport (HEBT) line at HIMAC. Beam test using these monitors have been carried out since April of 1995 to investigate a change of vertical beam size in synchrotron and a possibility of observing beam with high energy by one pass. In this paper the measurement system is mainly reported, and the preliminary results are also briefly presented. (author)

  12. Recent Trends in Electromagnetic Non-Destructive Sensing

    Directory of Open Access Journals (Sweden)

    Klara Capova

    2008-01-01

    Full Text Available The paper deals with material electromagnetic non-destructive testing (eNDT with emphasize on eddy current testing (ECT. Various modifications of ECT sensing are compared and discussed from the desired detected signal characteristics point of view. Except of the optimization of usual probe coils arrangements for the concrete applications, the new magnetic sensors as giant magneto-resistance (GMR and spin dependent tunneling (SDT are presented. The advanced ECT sensors are characterized by their sensitivity, frequency range and sensor dimensions

  13. Augmented reality application for industrial non-destructive inspection training

    Science.gov (United States)

    Amza, Catalin Gheorghe; Zapciu, Aurelian; Teodorescu, Octav

    2018-02-01

    Such a technology - Augmented Reality (AR) has great potential of use, especially for training purposes of new operators on using expensive equipment. In this context, the paper presents an augmented reality training system developed for phased-array ultrasonic non-destructive testing (NDT) equipment. The application has been developed using Unity 5.6.0 game-engine platform integrated with Vuforia sdk toolkit for devices with Android operating system. The test results performed by several NDT operators showed good results, thus proving the potential of using the application in the industrial field.

  14. 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.)

  15. 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.

  16. Laser active thermography for non-destructive testing

    International Nuclear Information System (INIS)

    Semerok, A.; Grisolia, C.; Fomichev, S.V.; Thro, P.Y.

    2013-01-01

    Thermography methods have found their applications in different fields of human activity. The non-destructive feature of these methods along with the additional advantage by automated remote control and tests of nuclear installations without personnel attendance in the contaminated zone are of particular interest. Laser active pyrometry and laser lock-in thermography for in situ non-destructive characterization of micrometric layers on graphite substrates from European tokamaks were under extensive experimental and theoretical studies in CEA (France). The studies were aimed to obtain layer characterization with cross-checking the layer thermal contact coefficients determined by active laser pyrometry and lock-in thermography. The experimental installation comprised a Nd-YAG pulsed repetition rate laser (1 Hz - 10 kHz repetition rate frequency, homogeneous spot) and a home-made pyrometer system based on two pyrometers for the temperature measurements in 500 - 2600 K range. For both methods, the layer characterization was provided by the best fit of the experimental results and simulations. The layer thermal contact coefficients determined by both methods were quite comparable. Though there was no gain in the measurements accuracy, lock-in measurements have proved their advantage as being much more rapid. The obtained experimental and theoretical results are presented. Some practical applications and possible improvements of the methods are discussed. (authors)

  17. Local defect resonance for sensitive non-destructive testing

    Science.gov (United States)

    Adebahr, W.; Solodov, I.; Rahammer, M.; Gulnizkij, N.; Kreutzbruck, M.

    2016-02-01

    Ultrasonic wave-defect interaction is a background of ultrasound activated techniques for imaging and non-destructive testing (NDT) of materials and industrial components. The interaction, primarily, results in acoustic response of a defect which provides attenuation and scattering of ultrasound used as an indicator of defects in conventional ultrasonic NDT. The derivative ultrasonic-induced effects include e.g. nonlinear, thermal, acousto-optic, etc. responses also applied for NDT and defect imaging. These secondary effects are normally relatively inefficient so that the corresponding NDT techniques require an elevated acoustic power and stand out from conventional ultrasonic NDT counterparts for their specific instrumentation particularly adapted to high-power ultrasonic. In this paper, a consistent way to enhance ultrasonic, optical and thermal defect responses and thus to reduce an ultrasonic power required is suggested by using selective ultrasonic activation of defects based on the concept of local defect resonance (LDR). A strong increase in vibration amplitude at LDR enables to reliably detect and visualize the defect as soon as the driving ultrasonic frequency is matched to the LDR frequency. This also provides a high frequency selectivity of the LDR-based imaging, i.e. an opportunity of detecting a certain defect among a multitude of other defects in material. Some examples are shown how to use LDR in non-destructive testing techniques, like vibrometry, ultrasonic thermography and shearography in order to enhance the sensitivity of defect visualization.

  18. Training guidelines in non-destructive testing techniques

    International Nuclear Information System (INIS)

    1987-02-01

    With the conclusion in 1979 of a successful Agency executed UNDP project in Argentina, whose aim was the establishment of a national non-destructive testing centre, the Agency was asked by other countries in the Latin American and Caribbean Region to evaluate the possibility of transferring this success to the whole region. In 1982, with the financial cooperation of UNFSSTD and UNIDO, a regional project was started with the principal objective of assisting the countries in the region to reinforce autonomous NDT capability through regional cooperation. One essential component of this project has been the harmonization of training through the development and use of course syllabi by the 17 countries now participating in the project. To this end, a Regional Working Group was formed and one of its tasks is the development of these syllabi for the more common NDT methods. This publication is a collection of the training programmes elaborated to date which have so far been followed by some 10,000 persons in the region who have received training in NDT as a direct result of the project. These syllabi take into account the development work done by the International Committee for Non-destructive Testing and many national training programmes, and are meant to be an objective guide to assist in the formation of NDT personnel

  19. Laser active thermography for non-destructive testing

    Science.gov (United States)

    Semerok, A.; Grisolia, C.; Fomichev, S. V.; Thro, P.-Y.

    2013-11-01

    Thermography methods have found their applications in different fields of human activity. The non-destructive feature of these methods along with the additional advantage by automated remote control and tests of nuclear installations without personnel attendance in the contaminated zone are of particular interest. Laser active pyrometry and laser lock-in thermography for in situ non-destructive characterization of micrometric layers on graphite substrates from European tokamaks were under extensive experimental and theoretical studies in CEA (France). The studies were aimed to obtain layer characterization with cross-checking the layer thermal contact coefficients determined by active laser pyrometry and lock-in thermography. The experimental installation comprised a Nd-YAG pulsed repetition rate laser (1 Hz - 10 kHz repetition rate frequency, homogeneous spot) and a home-made pyrometer system based on two pyrometers for the temperature measurements in 500 - 2600 K range. For both methods, the layer characterization was provided by the best fit of the experimental results and simulations. The layer thermal contact coefficients determined by both methods were quite comparable. Though there was no gain in the measurements accuracy, lock-in measurements have proved their advantage as being much more rapid. The obtained experimental and theoretical results are presented. Some practical applications and possible improvements of the methods are discussed.

  20. Non destructive testing of works of art by terahertz analysis

    Science.gov (United States)

    Bodnar, Jean-Luc; Metayer, Jean-Jacques; Mouhoubi, Kamel; Detalle, Vincent

    2013-11-01

    Improvements in technologies and the growing security needs in airport terminals lead to the development of non destructive testing devices using terahertz waves. Indeed, these waves have the advantage of being, on one hand, relatively penetrating. They also have the asset of not being ionizing. It is thus potentially an interesting contribution in the non destructive testing field. With the help of the VISIOM Company, the possibilities of this new industrial analysis method in assisting the restoration of works of art were then approached. The results obtained within this framework are presented here and compared with those obtained by infrared thermography. The results obtained show first that the THZ method, like the stimulated infrared thermography allows the detection of delamination located in murals paintings or in marquetries. They show then that the THZ method seems to allow detecting defects located relatively deeply (10 mm) and defects potentially concealed by other defects. It is an advantage compared to the stimulated infra-red thermography which does not make it possible to obtain these results. Furthermore, they show that the method does not seem sensitive to the various pigments constituting the pictorial layer, to the presence of a layer of "Japan paper" and to the presence of a layer of whitewash. It is not the case of the stimulated infrared thermography. It is another advantage of the THZ method. Finally, they show that the THZ method is limited in the detection of low-size defects. It is a disadvantage compared to the stimulated infrared thermography.

  1. Challenges of Non-Destructive Assay Waste Measurement

    International Nuclear Information System (INIS)

    Shull, A.H.

    2003-01-01

    Historically, the Savannah River Site (SRS) routinely produced special nuclear material (SNM), which provided stable measurement conditions for the non-destructive assay (NDA) methods. However, the main mission of SRS has changed from the production of SNM to the processing of waste and material stabilization. Currently, the purpose of processing is to recover the SNM from the waste and stabilization materials, much of which is from other DOE facilities. These missions are usually of a short duration, but require non-destructive assay (NDA) accountability measurements on materials of varying composition and geometric configuration. These missions usually have cost and time constraints, which sometimes require re-application of existing NDA methods to waste measurements. Usually, each new material or re-application of the NDA method to a different SNM campaign requires new standards and timely re-calibration of the method. These constraints provide numerous challenges for the NDA methods, particularly in the area of measurement uncertainty. This paper will discuss the challenges of these situations, mainly from a measurement and statistical point of view and provide some possible solutions to the problems encountered. Specific examples will be discussed for the segmented gamma scanner (SGS), neutron multiplicity counter (NMC) and passive neutron coincidence counter (PNCC), which are some of the most common NDA instruments at SRS

  2. Probability of Criticality for MOX SNF

    International Nuclear Information System (INIS)

    P. Gottlieb

    1999-01-01

    The purpose of this calculation is to provide a conservative (upper bound) estimate of the probability of criticality for mixed oxide (MOX) spent nuclear fuel (SNF)