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Sample records for mox47 fuel subjected

  1. 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.

  2. 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)

  3. 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

  4. 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)

  5. 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

  6. 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)

  7. 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)

  8. 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)

  9. 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.)

  10. 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

  11. 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)

  12. 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)

  13. 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)

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. 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)

  19. 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)

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. 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)

  5. 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)

  6. 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.)

  7. 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)

  8. 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)

  9. 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)

  10. 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)

  11. 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.

  12. 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

  13. 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)

  14. 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.)

  15. 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

  16. 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.)

  17. 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.)

  18. 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.

  19. 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)

  20. 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)

  1. 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)

  2. 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

  3. 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)

  4. 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)

  5. 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)

  6. 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

  7. 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)

  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. 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)

  10. 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.

  11. 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

  12. 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

  13. 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)

  14. 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)

  15. 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.)

  16. 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)

  17. 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)

  18. 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)

  19. 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.

  20. 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

  1. 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

  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. 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

  4. 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.

  5. 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

  6. 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

  7. 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.)

  8. 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)

  9. 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)

  10. 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

  11. 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)

  12. 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

  13. 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)

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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.

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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

  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. 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)

  6. 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)

  7. 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)

  8. 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

  9. 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

  10. 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.)

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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)

  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. 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.

  18. 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.

  19. 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)

  20. 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.

  1. 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

  2. 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)

  3. 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)

  4. 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)

  5. 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)

  6. 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

  7. 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)

  8. 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)

  9. 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)

  10. 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

  11. 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

  12. 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)

  13. 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)

  14. 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)

  15. 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)

  16. 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

  17. 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

  18. 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.

  19. 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)

  20. 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

  1. 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

  2. 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.

  3. 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)

  4. 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)

  5. 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).

  6. 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)

  7. 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

  8. 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.

  9. 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)

  10. 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

  11. 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)

  12. 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.

  13. 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

  14. 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

  15. 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)

  16. 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)

  17. 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.

  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. 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

  20. 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)

  1. 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)

  2. 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)

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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)

  8. 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

  9. 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)

  10. 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)

  11. 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)

  12. 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)

  13. 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)

  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; 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.

  15. 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)

  16. 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)

  17. 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)

  18. 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

  19. 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)

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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)

  5. 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)

  6. 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

  7. 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

  8. 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.

  9. 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)

  10. 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)

  11. 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

  12. 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.

  13. 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

  14. 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

  15. 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)

  16. 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)

  17. 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

  18. 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.

  19. 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

  20. 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)

  1. 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)

  2. 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)

  3. 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

  4. 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)

  5. 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.

  6. 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)

  7. 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)

  8. 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)

  9. 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)

  10. 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)

  11. 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

  12. 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)

  13. 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

  14. 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.

  15. 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

  16. 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)

  17. 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)

  18. Investigation of research and development subjects for the Very High Burnup Fuel

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Amano, Hidetoshi; Suzuki, Yasufumi; Furuta, Teruo; Nagase, Fumihisa; Suzuki, Masahide

    1993-06-01

    A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data and knowledge required for design and assessment of the fuel pellet, including the information on properties and irradiation performance of the TRU-added fuel. Finally, the R and D subjects were summarized, and a proposal was made on the way of development of the fuel pellet and cladding materials. (author)

  19. 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)

  20. 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.

  1. 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

  2. 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)

  3. 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

  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. 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. 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

  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. On the thermal evolution of Pu-rich agglomerates in MOX

    International Nuclear Information System (INIS)

    Verwerft, M.; Leenaers, A.; Lippens, M.; Mertens, L.

    1999-01-01

    From the experience accumulated so far on irradiated MOX fuel, its overall behaviour under irradiation is generally well predicted by existing fuel models. It appears however that additional data are still welcome to properly benchmark fission gas release models, mainly at elevated burnup. To this aim, an international research project, FIGARO, was initiated. Its goal was to provide thermal and fission gas release data og MOX at high burnup. Two MOX fuel rods irradiated to high burnup (50 GWd/tM peak pellet) but at lower power (less than 200 W/cm) were selected for segmentation and instrumentation with central thermocouple and pressure gauge. The instrumented segments were subjected to irradiations at variable linear power in the HALDEN MTR. Both temperature and internal pressure were online monitored during the ramp test. Afterwards, the rod segments were transported and extensively investigated. The paper focuses on the investigation of the evolution of the microstructure of Pu-rich agglomerates as a function of temperature

  9. 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)

  10. 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

  11. 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)

  12. 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

  13. 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

  14. 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

  15. Criticality safety philosophy for the Sellafield MOX plant

    International Nuclear Information System (INIS)

    Edge, Jane; Gulliford, Jim

    2003-01-01

    The Sellafield MOX Plant (SMP) has been operational since 2001, blending plutonium dioxide from THORP reprocessing operations, with uranium dioxide to produce Mixed Oxide (MOX) fuel elements. In handling the quantities of fuel associated with a commercial fuel fabrication plant, it is necessary to impose criticality controls. Plutonium dioxide (PuO 2 ), uranium dioxide (UO 2 ) and recycled MOX are mixed together in batches. An Engineered Protection System (EPS) prevents the production of MOX powder in excess of 20w/o Pu(fissile)/(Pu+U), achieved through the combination of a weight-based' system and a diverse 'neutron monitoring' radiometric system. The 'neutron monitoring' component of the EPS determines the fissile enrichment of the batch of MOX powder, based on pessimistic isotopic requirements of the PuO 2 feedstock powder. Guaranteeing the maximum MOX enrichment of 20w/o Pu(fissile)/(Pu + U) at an early stage of the fuel manufacturing process enables the criticality safety assessor to demonstrate that normal operations are deterministically safe. This paper describes in detail the EPS at the front end of plant and the engineered and operational protection in downstream areas. In addition plant operational experience in producing the first fuel assemblies is discussed. (author)

  16. 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.)

  17. 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)

  18. 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)

  19. 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

  20. 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 ...

  1. 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)

  2. 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)

  3. 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

  4. 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)

  5. 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)

  6. 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

  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. 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

  9. 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

  10. 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)

  11. 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)

  12. 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)

  13. 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

  14. 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)

  15. 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

  16. 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.

  17. 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.

  18. 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

  19. 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.

  20. 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.

  1. 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.

  2. Safeguards on MOX assemblies at LWRs

    International Nuclear Information System (INIS)

    Arenas Carrasco, J.; Koulikov, I.; Heinonen, O.J.; Arlt, R.; Grigoleit, K.; Clarke, R.; Swinhoe, M.

    2000-01-01

    Operating within the framework of the New Partnership Approach (NPA) for unirradiated MOX fuel assemblies in LWRs, the IAEA and EURATOM have gained experience in safeguarding 13 LWRs licensed to operate with MOX assemblies. In order to fulfil SIR requirements, verification methods and techniques capable of measuring MOX assemblies under water have been and are still being developed. These encompass both qualitative tests for the detection of plutonium (gross attribute tests) and quantitative tests for the measurement of the amount of plutonium (partial defect tests) and are based on gamma and neutron detection techniques. There are nine PWR and two BWR where the reactor and the spent fuel pond can be covered by the same surveillance device. These are Type I reactors where the reactor and the pond are located in the same hall. In these types of facilities relying on surveillance during the MOX refuelling is especially difficult at the BWRs due to the depth of the core pond. There are two PWR type facilities where the reactor and the spent fuel pond are located in different halls and cannot be covered by the same surveillance device (Type II). An open core camera has not been installed during refuelling and therefore indirect surveillance is currently used to survey MOX loading. Improvements are therefore required and are under consideration. After receipt at the facility, there are a few facilities which must keep the received fresh MOX fuel in wet storage, not only for a short period prior to refuelling, but for more than a year, until the next refuelling campaign. In these cases timely inspections for direct use fresh nuclear material require considerable inspection effort. Additionally, where human surveillance of core loading and finally core closure are necessary there is also a large demand for manpower. Either an agreement should be reached with the operators to delay the MOX loading until the end of the fuelling campaign, or alternative approaches should be

  3. 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)

  4. 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

  5. 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)

  6. 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)

  7. 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.

  8. 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)

  9. 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

  10. Recycling schemes of Americium targets in PWR/MOX cores

    International Nuclear Information System (INIS)

    Maldague, Th.; Pilate, S.; Renard, A.; Harislur, A.; Mouney, H.; Rome, M.

    1999-01-01

    From the orientation studies performed so far, both ways to recycle Am in PWR/MOX cores, homogeneous in MOX or heterogeneous in target pins, appear feasible, provided that enriched UO 2 is used as support of the MOX fuel. Multiple recycling can then proceed and stabilize Pu and Am quantities. With respect to the Pu multiple recycling strategy, recycling Am in addition needs 1/3 more 235 U, and creates 3 times more Curium. Thus, although feasible, such a fuel cycle is complicated and brings about a significant cost penalty, not quantified yet. The advantage of the heterogeneous option is to allow to manage in different ways the Pu in MOX fuel and the Am in target pins. For example, should Am remain combined to Cm after reprocessing, the recycling of a mix of Am+Cm could be deferred to let Cm transform into Pu before irradiation. The Am+Cm targets could also stay longer in the reactor, so as to avoid further reprocessing if possible. (author)

  11. 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)

  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. Evaluation of full MOX core capability for a 900 MWe PWR

    International Nuclear Information System (INIS)

    Joo, Hyung-Kook; Kim, Young-Jin; Jung, Hyung-Guk; Kim, Young-Il; Sohn, Dong-Seong

    1996-01-01

    Full MOX capability of a PWR core with 900 MWe capacity has been evaluated in view of plutonium consumption and design feasibility as an effective means for spent fuel management. Three full MOX cores have been conceptually designed; for annual cycle, for 18-month cycle, and for 18-month cycle with high moderation lattice. Fissile and total plutonium quantities at discharge are significantly reduced to 60% and 70% respectively of initial value for standard full MOX cores. It is estimated that one full MOX core demands about 1 tonne of plutonium per year to be reloaded, which is equivalent to reprocessing of spent nuclear fuels discharged from five nuclear reactors operating with uranium fuels. With low-leakage loading scheme, a full MOX core with either annual or 18-month cycle can be designed satisfactorily by installing additional rod cluster control system and modifying soluble boron system. Overall high moderation lattice case promises better core nuclear characteristics. (author)

  14. 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.

  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. 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

  17. Approach to customer qualification of the BNFL Sellafield Mox Plant

    International Nuclear Information System (INIS)

    Sullivan, P.

    2003-01-01

    BNFL started plutonium commissioning of its Sellafield MOX Plant (SMP) in December 2001, with the first MOX pellets being produced in May 2002. SMP was designed to manufacture a range of both PWR and BWR fuel types for a number of different customers. During commissioning and early MOX fuel manufacturing BNFL has been demonstrating its ability to both automatically manufacture and inspect MOX fuel to meet the requirements of different customers' specifications and fuel types. The qualification project consisted of common and project specific qualification. Common qualification was carried out to demonstrate BNFL could meet several customers' requirements during the same qualification test. Project specific qualification was carried out for one customer only as the fabrication or inspection equipment was specific to their fuel type. An example is the fuel assembly process. The reasons for BNFL carrying out common qualification were: - Develop a common qualified process to meet different customer specifications. - Minimise future qualifications prior to starting future fuel campaigns. - Ensure BNFL understands and effectively manages different customer requirements in SMP. BNFL has approached qualification of SMP systematically. Firstly the inspection system was qualified, and once completed the inspection system was then used in the qualification of the manufacturing process. (orig.)

  18. 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

  19. 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.

  20. 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)

  1. 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)

  2. 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

  3. Current developments of fuel fabrication technologies at the plutonium fuel production facility, PFPF

    International Nuclear Information System (INIS)

    Asakura, K.; Aono, S.; Yamaguchi, T.; Deguchi, M.

    2000-01-01

    The Japan Nuclear Cycle Development Institute, JNC, designed, constructed and has operated the Plutonium Fuel Production Facility, PFPF, at the JNC Tokai Works to supply MOX fuels to the proto-type Fast Breeder Reactor, FBR, 'MONJU' and the experimental FBR 'JOYO' with 5 tonMOX/year of fabrication capability. Reduction of personal radiation exposure to a large amount of plutonium is one of the most important subjects in the development of MOX fabrication facility on a large scale. As the solution of this issue, the PFPF has introduced automated and/or remote controlled equipment in conjunction with computer controlled operation scheme. The PFPF started its operation in 1988 with JOYO reload fuel fabrication and has demonstrated MOX fuel fabrication on a large scale through JOYO and MONJU fuel fabrication for this decade. Through these operations, it has become obvious that several numbers of equipment initially installed in the PFPF need improvements in their performance and maintenance for commercial utilization of plutonium in the future. Furthermore, fuel fabrication of low density MOX pellets adopted in the MONJU fuel required a complete inspection because of difficulties in pellet fabrication compared with high density pellet for JOYO. This paper describes new pressing equipment with a powder recovery system, and pellet finishing and inspection equipment which has multiple functions, such as grinding measurements of outer diameter and density, and inspection of appearance to improve efficiency in the pellet finishing and inspection steps. Another development of technology concerning an annular pellet and an innovative process for MOX fuel fabrication are also described in this paper. (author)

  4. 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)

  5. ZZ WPPR-FR-MOX/BNCMK, Benchmark on Pu Burner Fast Reactor

    International Nuclear Information System (INIS)

    Garnier, J.C.; Ikegami, T.

    1993-01-01

    Description of program or function: In order to intercompare the characteristics of the different reactors considered for Pu recycling, in terms of neutron economy, minor actinide production, uranium content versus Pu burning, the NSC Working Party on Physics of Plutonium Recycling (WPPR) is setting up several benchmark studies. They cover in particular the case of the evolution of the Pu quality and Pu fissile content for Pu recycling in PWRs; the void coefficient in PWRs partly fuelled with MOX versus Pu content; the physics characteristics of non-standard fast reactors with breeding ratios around 0.5. The following benchmarks are considered here: - Fast reactors: Pu Burner MOX fuel, Pu Burner metal fuel; - PWRs: MOX recycling (bad quality Pu), Multiple MOX recycling

  6. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor; Diseno de una recarga mixta con ensambles MOX de mayor relacion de moderacion para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

    2004-07-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  7. 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

  8. 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.

  9. 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.

  10. 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

  11. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    2013-01-01

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO 2 and MOX fuel rods, (3) analysis of isotopic composition data for UO 2 and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  12. Validation study of core analysis methods for full MOX BWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO{sub 2} and MOX fuel rods, (3) analysis of isotopic composition data for UO{sub 2} and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  13. 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)

  14. 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.)

  15. 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)

  16. 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)

  17. 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

  18. 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.

  19. 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)

  20. 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)

  1. VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4 - Revised Report

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, RJ

    2001-06-01

    The Task Force on Reactor-Based Plutonium Disposition (TFRPD) was formed by the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) to study reactor physics, fuel performance, and fuel cycle issues related to the disposition of weapons-grade (WG) plutonium as mixed-oxide (MOX) reactor fuel. To advance the goals of the TFRPD, 10 countries and 12 institutions participated in a major TFRPD activity: a blind benchmark study to compare code calculations to experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At Oak Ridge National Laboratory, the HELIOS-1.4 code system was used to perform the comprehensive study of pin-cell and MOX core calculations for the VENUS-2 MOX core benchmark study.

  2. Safety and licensing of MOX versus UO2 for BWRs and PWRs: Aspects applicable for civilian and weapons grade Pu

    International Nuclear Information System (INIS)

    Goldstein, L.; Malone, J.

    2000-01-01

    This paper reviews the safety and licensing differences between MOX and UO 2 BWR and PWR cores. MOX produced from the normal recycle route and from weapons grade material are considered. Reload quantities of recycle MOX assemblies have been licensed and continue to operate safely in European LWRs. In general, the European MOX assemblies in a reload are 2 . These studies indicated that no important technical or safety related issues have evolved from these studies. The general specifications used by fuel vendors for recycled MOX fuel and core designs are as follows: MOX assemblies should be designed to minimize or eliminate local power peaking mismatches with co-resident and adjacently loaded UO 2 assemblies. Power peaking at the interfaces arises from different neutronic behavior between UO 2 and MOX assemblies. A MOX core (MOX and UO 2 or all-MOX assemblies) should provide cycle energy equivalent to that of an all-UO 2 core. This applies, in particular, to recycle MOX applications. An important consideration when burning weapons grade material is rapid disposition which may not necessarily allow for cycle energy equivalence. The reactivity coefficients, kinetics data, power peaking, and the worth of shutdown systems with MOX fuel and cores must be such to meet the design criteria and fulfill requirements for safe reactor operation. Both recycle and weapons grade plutonium are considered, and positive and negative impacts are given. The paper contrasts MOX versus UO 2 with respect to safety evaluations. The consequences of some transients/accidents are compared for both types of MOX and UO 2 fuel. (author)

  3. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor

    International Nuclear Information System (INIS)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J.

    2004-01-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  4. 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

  5. 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)

  6. Development on nuclear fuel cycle business in Japan

    International Nuclear Information System (INIS)

    Usami, Kogo

    2002-01-01

    The Japan Nuclear Fuel Co., Ltd. (JNF) develops five businesses on nuclear fuel cycle such as uranium concentration, storage and administration of high level radioactive wastes, disposition of low level radioactive wastes, used fuel reprocessing, MOX fuel, at Rokkasho-mura in Aomori prefecture. Here were introduced on outline, construction and operation in reprocessing and MOX fuel works, outline, present state and future subjects on technical development of uranium concentration, outline and safety of disposition center on low level radioactive wastes, and storage and administration of high level radioactive wastes. (G.K.)

  7. 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)

  8. 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)

  9. 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

  10. Thermal property change of MOX and UO{sub 2} irradiated up to high burnup of 74 GWd/t

    Energy Technology Data Exchange (ETDEWEB)

    Nakae, Nobuo, E-mail: nakae-nobuo@jnes.go.jp [Japan Nuclear Energy Safety Organization (JNES), Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro [Japan Nuclear Energy Safety Organization (JNES), Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Kurematsu, Shigeru; Kosaka, Yuji [Nuclear Development Corporation (NDC), 622-12, Funaishikawa, Tokai-mura, Ibaraki 319-1111 (Japan); Yoshino, Aya; Kitagawa, Takaaki [Mitsubishi Nuclear Fuel Co., LTD. (MNF), 12-1, Yurakucho 1-Chome, Chiyoda-ku, Tokyo 100-0006 (Japan)

    2013-09-15

    Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO{sub 2} fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO{sub 2}. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO{sub 2} is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO{sub 2} at high burnup under the condition that the pellet–cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO{sub 2} before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO{sub 2}. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

  11. Study of advanced LWR cores for effective use of plutonium and MOX physics experiments

    International Nuclear Information System (INIS)

    Yamamoto, T.; Matsu-Ura, H.; Ueji, M.; Ota, H.; Kanagawa, T.; Sakurada, K.; Maruyama, H.

    1999-01-01

    Advanced technologies of full MOX cores have been studied to obtain higher Pu consumption based on the advanced light water reactors (APWRs and ABWRs). For this aim, basic core designs of high moderation lattice (H/HM ∼5) have been studied with reduced fuel diameters in fuel assemblies for APWRs and those of high moderation lattice (H/HM ∼6) with addition of extra water rods in fuel assemblies for ABWRs. The analysis of equilibrium cores shows that nuclear and thermal hydraulic parameters satisfy the design criteria and the Pu consumption rate increases about 20 %. An experimental program has been carried out to obtain the core parameters of high moderation MOX cores in the EOLE critical facility at the Cadarache Centre as a joint study of NUPEC, CEA and CEA's industrial partners. The experiments include a uranium homogeneous core, two MOX homogeneous cores of different moderation and a PWR assembly mock up core of MOX fuel with high moderation. The program was started from 1996 and will be completed in 2000. (author)

  12. KAERI results for BN600 full MOX benchmark (Phase 4)

    International Nuclear Information System (INIS)

    Lee, Kibog Lee

    2003-01-01

    The purpose of this document is to report the results of KAERI's calculation for the Phase-4 of BN-600 full MOX fueled core benchmark analyses according to the RCM report of IAEA CRP Action on U pdated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects. T he BN-600 full MOX core model is based on the specification in the document, F ull MOX Model (Phase4. doc ) . This document addresses the calculational methods employed in the benchmark analyses and benchmark results carried out by KAERI

  13. Effect of mixing state on criticality safety evaluation in MOX powder and additive

    International Nuclear Information System (INIS)

    Yamamoto, Toshihiro; Miyoshi, Yoshinori

    2005-01-01

    Criticality safety analyses are discussed in which MOX powder and additive (e.g. zinc-stearate) are mixed in a powder treatment process of MOX fuel fabrication. The multiplication factor k eff is largely affected by how they are mixed, i.e., how the density and volume change with the mixing. In general, k eff increases when MOX powder is mixed with zinc-stearate. However, plutonium content and density of MOX powder make a difference in the k eff 's changes. Especially, MOX powder with a higher plutonium content and a higher density is not always unsafe in terms of criticality if it is mixed with zinc-stearate. (author)

  14. Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

    International Nuclear Information System (INIS)

    Fridman, E.; Kliem, S.

    2011-01-01

    Research highlights: → Detailed 3D 100% Th-MOX PWR core design is developed. → Pu incineration increased by a factor of 2 as compared to a full MOX PWR core. → The core controllability under steady state conditions is demonstrated. - Abstract: Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated. Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin. The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core. Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities. The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B 4 C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution

  15. 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)

  16. High moderation MOX cores for effective use of plutonium in LWRs

    International Nuclear Information System (INIS)

    Hamamoto, Kazuko; Kanagawa, Takashi; Hiraiwa, Koji; Sakurada, Koichi; Moriwaki, Masanao; Aoyama, Motoo; Yamamoto, Toru; Ueji, Masao

    2001-01-01

    Conceptual design studies have been performed for high moderation full MOX cores aiming at increasing fissile Pu consumption rate (ratio of the consumed to the loaded fissile Pu) and reducing residual Pu in discharged MOX fuel. The BWR cores studied have hydrogen to heavy metal ratio(H/HM) of 5.9 with increasing water rods and 7.0 with reducing a fuel rod diameter based on a reference 9x9 fuel (H/HM=4.9) of ABWR. The PWR cores studied have H/HM of 5.0 and 6.0 with reducing a fuel rod diameter based on a reference 17x17 fuel (H/HM=4.0) of APWR. Equilibrium core design and plant safety analyses showed that those high moderation cores have compatibility with ABWR and APWR. The fissile Pu consumption rate is 22% larger than the full MOX cores with reference fuel of ABWR and 50% for APWR. The core performance and compatibility has been also evaluated in the condition of multi-recycle of Pu in these high moderation cores. Study has been conducted to evaluate the effect of introducing these high moderation cores in the fuel cycle of Japan. It shows that the high moderation cores produce 26% more cumulative electricity and reduce 22% stock of the fissile Pu by 2050 than the reference cores. (author)

  17. 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.

  18. 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

  19. 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) of the Westinghouse pressurized water reactor (PWR) design that has been proposed for use. with the Plutonium Disposition Program (Ref. 1, p. 2). This calculation uses a Monte Carlo technique similar to that used for ordinary commercial SNF (Ref. 2, Sections 2 and 5.2). Several scenarios, covering a range of parameters, are evaluated for criticality. Parameters specifying the loss of fission products and iron oxide from the waste package are particularly important. This calculation is associated with disposal of MOX SNF

  20. 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)

  1. Characteristics of MOX dissolution with silver mediated electrolytic oxidation method

    Energy Technology Data Exchange (ETDEWEB)

    Umeda, Miki; Nakazaki, Masato; Kida, Takashi; Sato, Kenji; Kato, Tadahito; Kihara, Takehiro; Sugikawa, Susumu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    MOX dissolution with silver mediated electrolytic oxidation method is to be applied to the preparation of plutonium nitrate solution to be used for criticality safety experiments at Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). Silver mediated electrolytic oxidation method uses the strong oxidisation ability of Ag(II) ion. This method is though to be effective for the dissolution of MOX, which is difficult to be dissolved with nitric acid. In this paper, the results of experiments on dissolution with 100 g of MOX are described. It was confirmed from the results that the MOX powder to be used at NUCEF was completely dissolved by silver mediated electrolytic oxidation method and that Pu(VI) ion in the obtained solution was reduced to tetravalent by means of NO{sub 2} purging. (author)

  2. Fission gas release of MOX with heterogeneous structure

    International Nuclear Information System (INIS)

    Nakae, N.; Akiyama, H.; Kamimura, K; Delville, R.; Jutier, F.; Verwerft, M.; Miura, H.; Baba, T.

    2015-01-01

    It is very useful for fuel integrity evaluation to accumulate knowledge base on fuel behavior of uranium and plutonium mixed oxide (MOX) fuel used in light water reactors (LWRs). Fission gas release is one of fuel behaviors which have an impact on fuel integrity evaluation. Fission gas release behavior of MOX fuels having heterogeneous structure is focused in this study. MOX fuel rods with a heterogeneous fuel microstructure were irradiated in Halden reactor (IFA-702) and the BR-3/BR-2 CALLISTO Loop (CHIPS program). The 85 Kr gamma spectrometry measurements were carried out in specific cycles in order to examine the concerned LHR (Linear Heat Rate) for fission gas release in the CHIPS program. The concerned LHR is defined in this paper to be the LHR at which a certain additional fission gas release thermally occurs. Post-irradiation examination was performed to understand the fission gas release behavior in connection with the pellet microstructure. The followings conclusions can be made from this study. First, the concerned LHR for fission gas release is estimated to be in the range of 20-23 kW/m with burnup over 37 GWd/tM. It is moreover guessed that the concerned LHR for fission gas release tends to decrease with increasing burnup. Secondly It is observed that FGR (fission gas release rate) is positively correlated with LHR when the LHR exceeds the concerned value. Thirdly, when burnup dependence of fission gas release is discussed, effective burnup should be taken into account. The effective burnup is defined as the burnup at which the LHR should be exceed the concerned value at the last time during all the irradiation period. And fourthly, it appears that FGR inside Pu spots is higher than outside and that retained (not released) fission gases mainly exist in the fission gas bubbles. Since fission gases in bubbles are considered to be easily released during fuel temperature increase, this information is very important to estimate fission gas release behavior

  3. Linear thermal expansion, thermal diffusivity and melting temperature of Am-MOX and Np-MOX

    International Nuclear Information System (INIS)

    Prieur, D.; Belin, R.C.; Manara, D.; Staicu, D.; Richaud, J.-C.; Vigier, J.-F.; Scheinost, A.C.; Somers, J.; Martin, P.

    2015-01-01

    Highlights: • The thermal properties of Np- and Am-MOX solid solutions were investigated. • Np- and Am-MOX solid solutions exhibit the same linear thermal expansion. • The thermal conductivity of Am-MOX is about 10% higher than that of Np-MOX. • The melting temperatures of Np-MOX and Am-MOX are 3020 ± 30 K and 3005 ± 30 K, respectively. - Abstract: The thermal properties of Np- and Am-MOX solid solution materials were investigated. Their linear thermal expansion, determined using high temperature X-ray diffraction from room temperature to 1973 K showed no significant difference between the Np and the Am doped MOX. The thermal conductivity of the Am-MOX is about 10% higher than that of Np-MOX. The melting temperatures of Np-MOX and Am-MOX, measured using a laser heating self crucible arrangement were 3020 ± 30 K and 3005 ± 30 K, respectively

  4. 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

  5. The MELOX MOX fabrication facility: history of an industrial success and future prospects

    International Nuclear Information System (INIS)

    Arslan, M.; Jacquet, R.; Krellmann, J.

    2005-01-01

    Along with the La Hague reprocessing plant, MELOX is part of the two industrial facilities that ensure the closure of the nuclear fuel cycle in France. Since started up in 1995, MELOX has specialized into recycling separated plutonium recovered from reprocessing operations performed at La Hague on spent UO 2 fuel. Capitalizing on the unique know-how acquired through thirty years of plutonium-based fuel fabrication at the Cadarache plant, this subsidiary of AREVA group has quickly become a worldwide expert in the industrial process of fabricating MOX: a fuel blend comprised of both uranium and plutonium oxides that allows at safely exploiting the energetic potential of plutonium. In order to address the various factors responsible for this industrial breakthrough, we will first present an overview of MELOX's history in regards of the emergence of a global MOX market. The added-value provided through treatment and recycling operations on spent fuel will be further described in terms of waste volume and radiotoxicity reduction. The emphasis will then be put on the total quality management policy that is at the core of MELOX's corporate strategy. Because MELOX has succeeded in meeting both productivity requirements and stringent quality constraints, it has won confidence from its European and Japanese clients. With increased production capacity of diversified MOX designs, MELOX is demonstrating the industrial efficiency of a new concept of MOX plants that is inspiring large construction projects in Japan, the US, and Russia. (authors)

  6. 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)

  7. Sodium fast reactor: an asset for a PWR UOX/MOX fleet - 5327

    International Nuclear Information System (INIS)

    Tiphine, M.; Coquelet-Pascal, C.; Girieud, R.; Eschbach, R.; Chabert, C.; Grosman, R.

    2015-01-01

    Due to its low fissile content, Pu from spent MOX fuels is sometimes regarded as not recyclable in LWR. Based on the existing French nuclear infrastructure (La Hague reprocessing plant and MELOX MOX manufacturing plant), AREVA and CEA have evaluated the conditions of Pu multi recycling in a 100% LWR fleet. As France is currently supporting a Fast Reactor prototype project, scenario studies have also been conducted to evaluate the contribution of a 600 MWe SFR in the LWR fleet. These scenario studies consider a nuclear fleet composed of 8 PWR 900 MWe, with or without the contribution of a SFR, and aim at evaluating the following points: -) the feasibility of Pu multi-recycling in PWR; -) the impact on the spent fuels storage; -) the reduction of the stored separated Pu; -) the impact on waste management and final disposal. The studies have been conducted with the COSI6 code, developed by CEA Nuclear Energy Direction since 1985, that simulates the evolution over time of a nuclear power plants fleet and of its associated fuel cycle facilities and provides material flux and isotopic compositions at each point of the scenario. To multi-recycle Pu into LWR MOX and to ensure flexibility, different reprocessing strategies were evaluated by adjusting the reprocessing order, the choice of used fuel assemblies according to their burn-up and the UOX/MOX proportions. The improvement of the Pu fissile quality and the kinetic of Pu multi-recycling in SFR depending on the initial Pu quality were also evaluated and led to a reintroduction of Pu in PWR MOX after a single irradiation in SFR, still in dilution with Pu from UOX to maintain a sufficient fissile quality. (authors)

  8. Analysis of a MOX-UO2 interface by the method of characteristics

    International Nuclear Information System (INIS)

    Chetaine, A.; Erradi, L.; Sanchez, R.; Zmijarevic, I.; Aniel-Buchheit, S.

    2005-01-01

    In the last few years many studies have been done to improve the ability of core reactors (PWR and BWR) to burn Plutonium fuel, either in mixed UO 2 /MOX pattern or full MOX pattern. The analysis of a MOX-UO 2 interface with the method of characteristics has been carried out. Comparisons with Monte Carlo and collision-probability calculations show that our results are in good agreement with those obtained by reference methods and qualify the method of characteristic as a reliable technique for such calculations. (authors)

  9. 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)

  10. 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.

  11. Buildup of radioxenon isotopes in MOX-assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Gniffke, Thomas; Kirchner, Gerald [Carl Friedrich von Weizsaecker-Centre for Science and Peace Research, Hamburg (Germany)

    2015-07-01

    Radioxenon is the main tracer for detection of nuclear tests conducted underground under the verification regime of the Comprehensive Nuclear Test Ban Treaty (CTBT). Since radioxenon is emitted by civilian sources too, like commercial nuclear reactors, source discrimination is still an important issue. Inventory calculations are necessary to predict which xenon isotopic ratios are built up in a reactor and how they differ from those generated by a nuclear explosion. The screening line actually used by the CTBT Organization for source discrimination is based on calculations for uranium fuel of various enrichments used in pressurized water reactors (PWRs). The usage of different fuel, especially mixed U/Pu oxide (MOX) assemblies with reprocessed plutonium, may alter the radioxenon signature of civilian reactors. In this talk, calculations of the radioxenon buildup in a MOX-assembly used in a commercial PWR are presented. Implications for the CTBT verification regimes are discussed and open questions are addressed.

  12. Mixed U/Pu oxide fuel fabrication facility co-processed feed, pelletized fuel

    International Nuclear Information System (INIS)

    1978-09-01

    Two conceptual MOX fuel fabrication facilities are discussed in this study. The first facility in the main body of the report is for the fabrication of LWR uranium dioxide - plutonium dioxide (MOX) fuel using co-processed feed. The second facility in the addendum is for the fabrication of co-processed MOX fuel spiked with 60 Co. Both facilities produce pellet fuel. The spiked facility uses the same basic fabrication process as the conventional MOX plant but the fuel feed incorporates a high energy gamma emitter as a safeguard measure against diversion; additional shielding is added to protect personnel from radiation exposure, all operations are automated and remote, and normal maintenance is performed remotely. The report describes the fuel fabrication process and plant layout including scrap and waste processing; and maintenance, ventilation and safety measures

  13. WESTINGHOUSE 17X17 MOX PWR ASSEMBLY - WASTE PACKAGE CRITICALITY ANALYSIS (SCPB: N/A)

    International Nuclear Information System (INIS)

    J.W. Davis

    1996-01-01

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to compare the criticality potential of Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel with the Design Basis spent nuclear fuel (SNF) analyzed previously (Ref. 5.1, 5.2). The basis of comparison will be the conceptual design Multi-Purpose Canister (MPC) PWR waste package concepts. The objectives of this evaluation are to show that the criticality potential of the MOX fuel is equal to or lower than the DBF or, if necessary, indicate what additional measures are required to make it so

  14. 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.

  15. 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.

  16. MOX Lead Assembly Fabrication at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Geddes, R.L. [Westinghouse Savannah River Company, AIKEN, SC (United States); Spiker, D.L.; Poon, A.P.

    1997-12-01

    The U. S. Department of Energy (DOE) announced its intent to prepare an Environmental Impact Statement (EIS) under the National Environmental Policy Act (NEPA) on the disposition of the nations weapon-usable surplus plutonium.This EIS is tiered from the Storage and Disposition of Weapons-Usable Fissile Material Programmatic Environmental Impact Statement issued in December 1996,and the associated Record of Decision issued on January, 1997. The EIS will examine reasonable alternatives and potential environmental impacts for the proposed siting, construction, and operation of three types of facilities for plutonium disposition. The three types of facilities are: a pit disassembly and conversion facility, a facility to immobilize surplus plutonium in a glass or ceramic form for disposition, and a facility to fabricate plutonium oxide into mixed oxide (MOX) fuel.As an integral part of the surplus plutonium program, Oak Ridge National Laboratory (ORNL) was tasked by the DOE Office of Fissile Material Disposition(MD) as the technical lead to organize and evaluate existing facilities in the DOE complex which may meet MD`s need for a domestic MOX fuel fabrication demonstration facility. The Lead Assembly (LA) facility is to produce 1 MT of usable test fuel per year for three years. The Savannah River Site (SRS) as the only operating plutonium processing site in the DOE complex, proposes two options to carry out the fabrication of MOX fuel lead test assemblies: an all Category I facility option and a combined Category I and non-Category I facilities option.

  17. Fuel assemblies for use in nuclear reactors

    International Nuclear Information System (INIS)

    Mochida, Takaaki.

    1987-01-01

    Purpose: To increase the plutonium utilization amount and improve the uranium-saving effect in the fuel assemblies of PWR type reactor using mixed uranium-plutonium oxides. Constitution: MOX fuel rods comprising mixed plutonium-uranium oxides are disposed to the outer circumference of a fuel assembly and uranium fuel rods only composed of uranium oxides are disposed to the central portion thereof. In such a fuel assembly, since the uranium fuel rods are present at the periphery of the control rod, the control rod worth is the same as that of the uranium fuel assembly in the prior art. Further, since about 25 % of the entire fuel rods is composed of the MOX fuel rods, the plutonium utilization amount is increased. Further, since the MOX fuel rods at low enrichment degree are present at the outer circumferential portion, mismatching at the boundary to the adjacent MOX fuel assembly is reduced and the problem of local power peaking increase in the MOX fuel assembly is neither present. (Kamimura, M.)

  18. Analysis of high moderation full MOX BWR core physics experiments BASALA

    International Nuclear Information System (INIS)

    Ishii, Kazuya; Ando, Yoshihira; Takada, Naoyuki; Kan, Taro; Sasagawa, Masaru; Kikuchi, Tsukasa; Yamamoto, Toru; Kanda, Ryoji; Umano, Takuya

    2005-01-01

    Nuclear Power Engineering Corporation (NUPEC) has performed conceptual design studies of high moderation full MOX LWR cores that aim for increasing fissile Pu consumption rate and reducing residual Pu in discharged MOX fuel. As part of these studies, NUPEC, French Atomic Energy Commission (CEA) and their industrial partners implemented an experimental program BASALA following MISTRAL. They were devoted to measuring the core physics parameters of such advanced cores. The MISTRAL program consists of one reference UO 2 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice. As for MISTRAL, the analysis results have already been reported on April 2003. The BASALA program consists of two high moderation full MOX BWR mock-up cores for operating and cold stand-by conditions. NUPEC has analyzed the experimental results of BASALA with the diffusion and the transport calculations by the SRAC code system and the continuous energy Monte Carlo calculations by the MVP code with the common nuclear data file, JENDL-3.2. The calculation results well reproduce the experimental data approximately within the same range of the experimental uncertainty. The analysis results of MISTRAL and BASALA indicate that these applied analysis methods have the same accuracy for the UO 2 and MOX cores, for the different moderation MOX cores, and for the homogeneous and the mock-up MOX cores. (author)

  19. 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)

  20. 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

  1. Development of simulation code for MOX dissolution using silver-mediated electrochemical method (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Kida, Takashi; Umeda, Miki; Sugikawa, Susumu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    MOX dissolution using silver-mediated electrochemical method will be employed for the preparation of plutonium nitrate solution in the criticality safety experiments in the Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). A simulation code for the MOX dissolution has been developed for the operating support. The present report describes the outline of the simulation code, a comparison with the experimental data and a parameter study on the MOX dissolution. The principle of this code is based on the Zundelevich's model for PuO{sub 2} dissolution using Ag(II). The influence of nitrous acid on the material balance of Ag(II) is taken into consideration and the surface area of MOX powder is evaluated by particle size distribution in this model. The comparison with experimental data was carried out to confirm the validity of this model. It was confirmed that the behavior of MOX dissolution could adequately be simulated using an appropriate MOX dissolution rate constant. It was found from the result of parameter studies that MOX particle size was major governing factor on the dissolution rate. (author)

  2. 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)

  3. 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

  4. 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.

  5. 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

  6. Overview of neutronic fuel assembly design and in-core fuel management

    International Nuclear Information System (INIS)

    Porsch, D.; Charlier, A.; Meier, G.; Mougniot, J.C.; Tsuda, K.

    2000-01-01

    The civil and military utilization of nuclear power results in stockpiles of spent fuel and separated plutonium. Recycling of the recovered plutonium in Light Water Reactors (LWR) is currently practiced in Belgium, France, Germany, and Switzerland, in Japan it is in preparation. Modern MOX fuel, with its optimized irradiation and reprocessing behavior, was introduced in 1981. Since then, about 1700 MOX fuel assemblies of different mechanical and neutronic design were irradiated in commercial LWRs and reached fuel assembly averaged exposures of up to 51.000 MWd/t HM. MOX fuel assemblies reloaded in PWR have an average fissile plutonium content of up to 4.8 w/o. For BWR, the average fissile plutonium content in actual reloads is 3.0 w/o. Targets for the MOX fuel assembly design are the compatibility to uranium fuel assemblies with respect to their mechanical fuel rod and fuel assembly design, they should have no impact on the flexibility of the reactor operation, and its reload should be economically feasible. In either cycle independent safety analyses or individually for each designed core it has to be demonstrated that recycling cores meet the same safety criteria as uranium cores. The safety criteria are determined for normal operation and for operational as well as design basis transients. Experience with realized MOX core loadings confirms the reliability of the applied modern design codes. Studies for reloads of advanced MOX assemblies in LWRs demonstrate the feasibility of a future development of the thermal plutonium recycling. New concepts for the utilization of plutonium are under consideration and reveal an attractive potential for further developments on the plutonium exploitation sector. (author)

  7. Safety problems related to the use of MOX assemblies in PWRS

    International Nuclear Information System (INIS)

    Gouffon, A.; Merle, J.P.

    1989-12-01

    Curtailment of the LMFBR program along with the satisfactory performance of the La Hague reprocessing plant, with the consequent availability of large quantities of plutonium, provides Electricite de France (EDF) with the possibility of burning mixed uranium and plutonium oxide fuel (MOX fuel) in the core of certain PWR power plant reactors, hence reducing enriched uranium fuel requirements. Design provision has in fact been made for this possibility on sixteen 900 MWe plant units and is explicitly authorized in the relevant authorization decrees. In this paper, we have restricted our discussion to safety aspects pertaining to utilization of the fuel in the reactor. Generally speaking, the Safety Analysis Department has checked that the provisions made by EDF and/or the scheduled plant modifications enabled reactor unit operating safety to be maintained at the same level as for standard fuel management systems and that, in particular, the recycling of 30% MOX assemblies was compatible with observance, under accident conditions, of the same safety criteria as for all uranium cores

  8. Nuclear Fuel Design Technology Development for the Future Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Cheon, Jin Sik; Oh, Je Yong; Yim, Jeong Sik; Sohn, Dong Seong; Lee, Byung Uk; Ko, Han Suk; So, Dong Sup; Koo, Dae Seo

    2006-04-15

    The test MOX fuels have been irradiated in the Halden reactor, and their burnup attained 40 GWd/t as of October 2005. The fuel temperature and internal pressure were measured by the sensors installed in the fuels and test rig. The COSMOS code, which was developed by KAERI, well predicted in-reactor behavior of MOX fuel. The COSMOS code was verified by OECD-NEA benchmarks, and the result confirmed the superiority of COSMOS code. MOX in-pile database (IFA-629.3, IFA-610.2 and 4) in Halden was also used for the verification of code. The COSMOS code was improved by introducing Graphic User Interface (GUI) and batch mode. The PCMI analysis module was developed and introduced by the new fission gas behavior model. The irradiation test performed under the arbitrary rod internal pressure could also be analyzed with the COSMOS code. Several presentations were made for the preparation to transfer MOX fuel performance analysis code to the industry, and the transfer of COSMOS code to the industry is being discussed. The user manual and COSMOS program (executive file) were provided for the industry to test the performance of COSMOS code. To envisage the direction of research, the MOX fuel research trend of foreign countries, specially focused on USA's GENP policy, was analyzed.

  9. 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

  10. 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.

  11. Thorium utilization as a Pu-burner: proposal of Plutonium-Thorium Mixed Oxide (PT-MOX) Project

    International Nuclear Information System (INIS)

    Aizawa, Otohiko

    2000-01-01

    In this paper, a Pu-Th mixed oxide (PT-MOX) project is proposed for a thorium utilization and a plutonium burning. None of plutonium can be newly produced from PT-MOX fuel, and the plutonium mass of about 1 ton can be consumed with one reactor (total heavy metal assumed: 100 tons) for 1 year. In order to consume plutonium produced from usual Light Water Reactor, it should be better to operate one PT-MOX reactor for three to five Light Water Reactors. (author)

  12. MELOX fuel fabrication plant: Operational feedback and future prospects

    International Nuclear Information System (INIS)

    Hugelmann, D.; Greneche, D.

    2000-01-01

    As of December 1, 1998, 32 Europeans LWRs are loaded with MOX fuel. It clearly means that plutonium recycling in MOX fuels is a mature industry, with successful operational experience in fabrication plants in some European countries, especially in France. Indeed, the recycling of plutonium generated in LWRs is one of the objectives of the full Reprocessing-Conditioning-Recycling (RCR) strategy chosen by France in the 70's. The most impressive results of this strategy, is the fact that 31 of the 32 reactors are loaded with MOX fuels supplied by the COGEMA Group from the same efficient fabrication process, the MIMAS process, improved for the MELOX plant to become the A-MIMAS process. In France, 17 reactors are already loaded and 11 additional reactors are technically suited to do so. Indeed, the EDF MOX program plans to use MOX in 28 of its 57 reactors. An EDF 900 MWe reactor core contains 157 assemblies of 264 rods each. 52 fuel assemblies per year are necessary for a 'UO 2 3-batches-MOX 3-batches' core management. In this case, a third of the UO 2 and a third of the MOX assemblies are replaced yearly, that means 36 UO 2 fuel assemblies and 16 MOX fuel assemblies. Some MOX fuelled reactors have now switched from the previously described core management to a so-called 'hybrid core management'. In this case, a quarter of UO 2 assemblies is replaced yearly. The first EDF reactor loaded with MOX fuel was Saint-Laurent B1, in 1987. The in-core experience, based on several hundred assemblies loaded, with reloading on a 1/3 cycle basis, shows that there is no operational difference between UO 2 and MOX fuels, both in terms of performance and safety. MOX fueling of 900 MWe EDF's PWRs, with a limited in-core MOX ratio of 30%, has needed only minor adaptations, such as addition of control rods, modification of the boron concentration in the cooling system and precaution against radiation exposure, easy to set up (optimisation of the fresh MOX fuel handling process, remote

  13. Reactor fuel element and fuel assembly

    International Nuclear Information System (INIS)

    Okada, Seiji; Ishida, Tsuyoshi; Ikeda, Atsuko.

    1997-01-01

    A mixture of fission products and burnable poisons is disposed at least to a portion between MOX pellets to form a burnable poison-incorporated fuel element without mixing burnable poisons to the MOX pellets. Alternatively, a mixture of materials other than the fission products and burnable poisons is formed into disks, a fuel lamination portion is divided into at least to two regions, and the ratio of number of the disks of the mixture relative to the volume of the region is increased toward the lower portion of the fuel lamination portion. With such a constitution, the axial power distribution of fuels can be made flat easily. Alternatively, the thickness of the disk of the mixture is increased toward the lower region of the fuel lamination portion to flatten the axial power distribution of the fuels in the same manner easily. The time and the cost required for the manufacture are reduced, and MOX fuels filled with burnable poisons with easy maintenance and control can be realized. (N.H.)

  14. 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

  15. Fuel cycle in Japanese Fugen - HWR

    International Nuclear Information System (INIS)

    1979-04-01

    This paper describes the use of plutonium-bearing fuel in the Japanese Fugen-HWR. The Fugen-HWR is a pressure tube type, boiling light water cooled, and heavy water moderated reactor, which by using plutonium fuel (MOX) achieves the advantage of high neutron economy. The characteristics of the reactor are discussed, particularly its ability to operate with several different types of fuel - Pu-natural U MOX, Pu-Depleted U (from spent LWR fuel) MOX, Pu-Depleted U (from enrichment tails) MOX, and enriched UO 2 . The natural U and separative work units saved are given and the fuel management and control of the reactor discussed. Non-proliferation and safety considerations are given. The Fugen-HWR achieved 100% power rating in the autumn of 1979

  16. Neutronics benchmark of a MOX assembly with near-weapons-grade plutonium

    International Nuclear Information System (INIS)

    Difilippo, F.C.; Fisher, S.E.

    1998-01-01

    One of the proposed ways to dispose of surplus weapons-grade plutonium (Pu) is to irradiate the high-fissile material in light-water reactors in order to reduce the Pu enrichment to the level of spent fuels from commercial reactors. Considerable experience has been accumulated about the behavior of mixed-oxide (MOX) uranium and plutonium fuels for plutonium recycling in commercial reactors, but the experience is related to Pu enrichments typical of spent fuels quite below the values of weapons-grade plutonium. Important decisions related to the kind of reactors to be used for the disposition of the plutonium are going to be based on calculations, so the validation of computational algorithms related to all aspects of the fuel cycle (power distributions, isotopics as function of the burnup, etc.), for weapons-grade isotopics is very important. Analysis of public domain data reveals that the cycle-2 irradiation in the Quad cities boiling-water reactor (BWR) is the most recent US destructive examination. This effort involved the irradiation of five MOX assemblies using 80 and 90% fissile plutonium. These benchmark data were gathered by General Electric under the sponsorship of the Electric Power Research Institute. It is emphasized, however, that global parameters are not the focus of this benchmark, since the five bundles containing MOX fuels did not significantly affect the overall core performance. However, since the primary objective of this work is to compare against measured post-irradiation assembly data, the term benchmark is applied here. One important reason for performing the benchmark on Quad Cities irradiation is that the fissile blends (up to 90%) are higher than reactor-grade and, quite close to, weapons-grade isotopics

  17. 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

  18. 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.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  20. 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

  1. The feasibility study on fuel types for the KALIMER

    International Nuclear Information System (INIS)

    Hwang, W.; Nam, C.; Yim, J. S.; Na, B. C.; Hahn, D. H.; Kim, Y. I.; Kim, Y. C.; Park, C. K.

    1997-08-01

    The economics of LMR is largely dependent on the construction cost of the power plant, and the fuel cycle options usually constitute 20 to 30 % of total electricity generation cost. The choice of fuel cycle technology and the fuel type is important in order to develop a LMR with better economics, performance and safety. The LMR fuel types, whose performances have been proven up to 15 at% burnup, are MOX and IFR metal fuel. The base alloy, binary (U-10% Zr) metal fuel with HT9 is used as structural materials of KALIMER. The design concept of KALIMER fuel has been established through the investigation of technical feasibilities on the fuel and recycle systems for MOX and IFR metal fuel. According to the results of comparative analysis for MOX and metal fuel, metal fuel is better than MOX in view of safety, in-reactor performance, nuclear characteristics, economics and non-proliferation, while MOX fuels have advantages in the developmental status and technical cooperation potential. The overall performance of binary (U-10% Zr) metal fuel with HT9 cladding, which is a potential start-up fuel for KALIMER, is not only superior to that of MOX fuel, but also has enough technical feasibility in its high-burnup performance, safety and economics. (author). 54 ref., 13 tabs., 20 figs

  2. 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

    The Spent Fuel Stability Project of the European Union aims to develop a model predicting the radionuclide release rate from spent fuel as a function of time for geological disposal conditions. In the first phase of the project, an important aspect consists of the model development focused on defining the instant release fraction (IRF), which represents the fraction of the inventory of safety-relevant radionuclides that may be rapidly released from the fuel and fuel assembly materials at the time of canister breaching. The locations of these preferentially released radionuclides, their quantities, the evidence for their release and proposed estimates of the IRF for the key safety-relevant radionuclides for the case of fuel shortly after discharge from the reactor are the subjects of the present report. Spent fuel assemblies comprise several materials, including uranium oxide, Zircaloy and various steels or nickel alloys used in the structural components of fuel assemblies. Information on the distribution of both activation products and fission products in all these materials must be taken into account in deriving IRF values. Information is presented on the radionuclide distributions in the various materials and IRF values for key radionuclides are proposed. The IRF increases with burnup, particularly above 50 GWd/t IHM . The estimated IRF values are functions of the numbers of spent fuel assemblies in various burnup intervals. Use of bounding fission gas release (FGR) values lead to overestimates of derived IRF values. The problem of uncertainties must be given attention because there is considerable scatter in FGR data, as well as in data from fission product leaching studies. The following approaches and definitions are adopted in this report: a) Best estimate, based on a good understanding of the mechanism and a good quality database; b) Bounding or pessimistic estimate based on data and process understanding that provides a maximum for the range of derived

  3. Decommissioning the Belgonucleaire Dessel MOX plant: presentation of the project and situation end august 2013

    Energy Technology Data Exchange (ETDEWEB)

    Cuchet, J.M. [TRACTEBEL ENGINEERING, Avenue Ariane, 7, B1200 Brussels (Belgium); Libon, H.; Verheyen, C. [BELGONUCLEAIRE S.A. / N.V. Europalaan, 20, B2480 Dessel (Belgium); Bily, J. [STUDSVIK GmbH, Karlsruher Strasse, 20, D75179 Pforzheim,(Germany); Boden, S. [SCK-CEN, Boeretang, 200, B2400 Mol (Belgium); Joffroy, F. [TECNUBEL N.V., Zandbergen, 1, B2480 Dessel (Belgium); Walthery, R. [BELGOPROCESS, Gravenstraat, 73, B2480 Dessel (Belgium)

    2013-07-01

    Belgonucleaire has been operating the Dessel MOX plant at an industrial scale between 1986 and 2006. During this period, 40 metric tons of plutonium (HM) have been processed into 90 reloads of MOX fuel for commercial light water reactors. The decision to stop the production in 2006 and to decommission the MOX plant was the result of the shrinkage of the MOX fuel market due to political and commercial factors. As a significant part of the decommissioning project of the Dessel MOX plant, about 170 medium-sized glove-boxes and about 1.200 metric tons of structure and equipment outside the glove-boxes are planned for dismantling. The license for the dismantling of the MOX plant was granted by Royal Decree in 2008 and the dismantling started in March 2009. The dismantling works are carried out by an integrated organization under leadership and responsibility of Belgonucleaire; this organization includes 3 main contractors, namely Tecnubel N.V., the THV ('Tijdelijke HandelsVereniging') Belgoprocess / SCK-CEN and Studsvik GmbH and Tractebel Engineering as project manager. In this paper, after having described the main characteristics of the project, the authors review the different organizational and technical options considered for the decommissioning of the glove-boxes; thereafter the main decision criteria (qualification of personnel and of processes, confinement, cutting techniques and radiation protection, safety aspects, alpha-bearing waste management) are analyzed as well. Finally the progress, the feedback and the lessons learned at the end of August 2013 are presented, giving the principal's and contractors point of view. (authors)

  4. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

    In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

  5. 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

  6. Neutronic and Logistic Proposal for Transmutation of Plutonium from Spent Nuclear Fuel as Mixed-Oxide Fuel in Existing Light Water Reactors

    International Nuclear Information System (INIS)

    Trellue, Holly R.

    2004-01-01

    The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium

  7. 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.)

  8. Disposition of excess plutonium using ''off-spec'' MOX pellets as a sintered ceramic waste form

    International Nuclear Information System (INIS)

    Armantrout, G.A.; Jardine, L.J.

    1996-02-01

    The authors describe a potential strategy for the disposition of excess weapons plutonium in a way that minimizes (1) technological risks, (2) implementation costs and completion schedules, and (3) requirements for constructing and operating new or duplicative Pu disposition facilities. This is accomplished by an optimized combination of (1) using existing nuclear power reactors to ''burn'' relatively pure excess Pu inventories as mixed oxide (MOX) fuel and (2) using the same MOX fuel fabrication facilities to fabricate contaminated or impure excess Pu inventories into an ''off-spec'' MOX solid ceramic waste form for geologic disposition. Diversion protection for the SCWF to meet the ''spent fuel standard'' introduced by the National Academy of Sciences can be achieved in at least three ways. (1) One can utilize the radiation field from defense high-level nuclear waste by first packaging the SCWF pellets in 2- to 4-L cans that are subsequently encapsulated in radioactive glass in the Defense Waste Processing Facility (DWPF) glass canisters (a ''can-in-canister'' approach). (2) One can add 137 Cs (recovered from defense wastes at Hanford and currently stored as CsCl in capsules) to an encapsulating matrix such as cement for the SCWF pellets in a small hot-cell facility and thus fabricate large monolithic forms. (3) The SCWF can be fabricated into reactor fuel-like pellets and placed in tubes similar to fuel assemblies, which can then be mixed in sealed repository containers with irradiated spent nuclear fuel for geologic disposition

  9. Computational benchmark on the void reactivity effect in MOX lattices. Contribution to a NEA-NSC benchmark study organized by the Working Party on Plutonium Recycling

    International Nuclear Information System (INIS)

    Freudenreich, W.E.; Aaldijk, J.K.

    1994-08-01

    The Working Party on Plutonium Recycling of the Nuclear Science Committee of the OECD Nuclear Energy Agency has initiated a benchmark study on the calculation of the void reactivity effect in MOX lattices. The results presented here were obtained with the continuous energy, generalized geometry Monte Carlo transport code MCNP. The cross-section libraries used were processed from the JEF-2.2 evaluation taking into account selfshielding in the unresolved resonance ranges (selfshielding in the resolved resonance ranges is treated by MCNP). For an infinite lattice of unit cells a positive void reactivity effect was found only for the MOX fuel with the largest Pu content. For an infinite lattice of macro cells (voidable inner zone with different fuel mixtures surrounded by an outer zone of UO 2 fuel with moderator) a positive void reactivity effect was obtained for the three MOX fuel types considered. The results are not representative for MOX-loaded power reactor lattices, but serve only to intercompare reactor physics codes and libraries. (orig.)

  10. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  11. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    International Nuclear Information System (INIS)

    Cowell, B.S.; Fisher, S.E.

    1999-01-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option

  12. 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)

  13. Effects of cooling time on a closed LWR fuel cycle

    International Nuclear Information System (INIS)

    Arnold, R. P.; Forsberg, C. W.; Shwageraus, E.

    2012-01-01

    In this study, the effects of cooling time prior to reprocessing spent LWR fuel has on the reactor physics characteristics of a PWR fully loaded with homogeneously mixed U-Pu or U-TRU oxide (MOX) fuel is examined. A reactor physics analysis was completed using the CASM04e code. A void reactivity feedback coefficient analysis was also completed for an infinite lattice of fresh fuel assemblies. Some useful conclusions can be made regarding the effect that cooling time prior to reprocessing spent LWR fuel has on a closed homogeneous MOX fuel cycle. The computational analysis shows that it is more neutronically efficient to reprocess cooled spent fuel into homogeneous MOX fuel rods earlier rather than later as the fissile fuel content decreases with time. Also, the number of spent fuel rods needed to fabricate one MOX fuel rod increases as cooling time increases. In the case of TRU MOX fuel, with time, there is an economic tradeoff between fuel handling difficulty and higher throughput of fuel to be reprocessed. The void coefficient analysis shows that the void coefficient becomes progressively more restrictive on fuel Pu content with increasing spent fuel cooling time before reprocessing. (authors)

  14. Development of a reference scheme for MOX lattice physics calculations

    International Nuclear Information System (INIS)

    Finck, P.J.; Stenberg, C.G.; Roy, R.

    1998-01-01

    The US program to dispose of weapons-grade Pu could involve the irradiation of mixed-oxide (MOX) fuel assemblies in commercial light water reactors. This will require licensing acceptance because of the modifications to the core safety characteristics. In particular, core neutronics will be significantly modified, thus making it necessary to validate the standard suites of neutronics codes for that particular application. Validation criteria are still unclear, but it seems reasonable to expect that the same level of accuracy will be expected for MOX as that which has been achieved for UO 2 . Commercial lattice physics codes are invariably claimed to be accurate for MOX analysis but often lack independent confirmation of their performance on a representative experimental database. Argonne National Laboratory (ANL) has started implementing a public domain suite of codes to provide for a capability to perform independent assessments of MOX core analyses. The DRAGON lattice code was chosen, and fine group ENDF/B-VI.04 and JEF-2.2 libraries have been developed. The objective of this work is to validate the DRAGON algorithms with respect to continuous-energy Monte Carlo for a suite of realistic UO 2 -MOX benchmark cases, with the aim of establishing a reference DRAGON scheme with a demonstrated high level of accuracy and no computing resource constraints. Using this scheme as a reference, future work will be devoted to obtaining simpler and less costly schemes that preserve accuracy as much as possible

  15. 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.

  16. Use of the program TNHXY in assemblies type MOX in comparison with CASMO-4; Utilizacion del programa TNHXY en ensambles tipo MOX en comparacion con CASMO-4

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J. V.; Enriquez C, P. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

    2011-11-15

    In this work a comparison is made in the analysis of fuel assemblies type MOX among the CASMO-4 code and the program TNHXY (Transport of neutrons with Hybrid Nodal schemes in X Y geometry) which solves the equation of neutrons transport in stationary state and X Y geometry using nodal schemes type finite element -hybrid-, such named in correspondence to the parameters that interpolate. The program TNHXY has been validated previously by means of different test problems or benchmark that some authors have solved using other numeric techniques. In addition to analyzing assemblies type BWR. Some of the codes with which have been realized the validations are TWOTRAN as well as other commercial codes as, Helios, MCNP-4B and Cpm-3. The reason of to do this comparative is to able to observe the versatility of the program TNHXY with regard to CASMO-4 relating to the assemblies analysis type MOX and BWR, offering an alternative in the analysis of the same assemblies and with this comparison is confirmed even more the program TNHXY. For the comparison was analyzed a fuel assembly of the type GNF2 for a reactor type BWR that contains MOX with 10 enrichment types for a specific burnt pass. (Author)

  17. 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

  18. FUJI - a comparative irradiation test with pellet, sphere-pac, and vipac fuel

    International Nuclear Information System (INIS)

    Hellwig, C.; Bakker, K.; Ozawa, T.; Nakamura, M.; Kihara, Y.

    2004-01-01

    Particle fuels such as sphere-pac and vipac fuels have been considered as promising fuel systems for fast reactors, due to their inherent potential in remote operation, cost reduction and incineration of minor actinides or low-decontaminated plutonium. The FUJI test addresses the questions of fabrication of MOX particle fuels with high Pu content (20%) and its irradiation behaviour during the start-up phase. Four kinds of fuel, i.e. MOX sphere-pac, MOX vipac, MOX pellet and Np-MOX sphere-pac fuel, have been and will be simultaneously irradiated under identical conditions in the High Flux Reactor in Petten. First results show that the particle fuel undergoes a dramatic structure change already at the very beginning of the irradiation when the maximum power is reached. The structural changes, i.e. the formation of a central void and the densification of fuel, decrease the fuel central temperature. Thus the fast and strong restructuring helps to prevent central fuel melting at high power levels. (authors)

  19. 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.

  20. A NOVEL APPROACH TO FIND OPTIMIZED NEUTRON ENERGY GROUP STRUCTURE IN MOX THERMAL LATTICES USING SWARM INTELLIGENCE

    Directory of Open Access Journals (Sweden)

    M. AKBARI

    2013-12-01

    Full Text Available Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that UO2–PUO2 (MOX is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the UO2 fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of H2O moderated UO2–PuO2 (MOX lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure.

  1. Fuel R and D international programmes, a way to demonstrate future fuel performances

    International Nuclear Information System (INIS)

    Vanderborck, Y.; Mertens, L.; Dekeyser, J.; Sannen, L.

    1997-01-01

    As a MOX fuel manufacturer, BELGONUCLEAIRE have spent more than 15 years promoting and managing International R and D Programmes, many of them in close cooperation with SCK''centrdot'' CEN. Such programmes dedicated to MOX versus UO 2 fuel behaviour are most of the time based on irradiation in research reactors in which the investigated fuel is submitted to power variations and to ramp testing or are performed in commercial reactors. This paper is focused on recent programmes concerned by high and medium burn-up in BWR and PWR conditions for MOX fuel. It will present also the new opportunities for new programmes. The goals, the programmes descriptions and the expected data being part of these R and D programmes is presented. (author)

  2. EUROFAB: fabrication of four MOX lead tests assemblies for the US DOE

    International Nuclear Information System (INIS)

    Jean-Pierre Bariteau

    2006-01-01

    In a multilateral agreement, the United States (US) and the Russian Federation agreed to reduce their respective weapons stockpiles by each country disposing of 34 tons of military origin plutonium. On behalf of the US government, the Department of Energy contracted with Duke, COGEMA, Stone and Webster (DCS) to design a Mixed Oxide Fuel Fabrication facility (MFFF) which would be built and operated at the DOE Savannah River Site near Aiken, South Carolina. This plant will transform the US excess weapons stockpile into MOX fuel, which will be used it in existing domestic commercial power reactors. The MFFF is based on a replication of AREVA existing facilities (La Hague for Pu polishing and Melox for MOX fabrication). In parallel with the design, construction and startup of the MFFF facility, DOE commissioned fabrication and irradiation of 4 lead test assemblies in one of the Mission Reactors to assist in obtaining NRC approval for MOX fuel loading in US NPPs prior to the production phase of the MFFF facility. This program was named 'EUROFAB', since fabrication had to be made in Europe because no facility implementing the MFFF technology was existing in the USA. The COGEMA Recycling Business unit transmitted a bid to DCS in April 2003, which proposed to perform Eurofab fabrication in its Cadarache (pellets and rods) and Melox (assembly mounting) facilities. In August 2003, the decision was made by DCS, on behalf of the DOE, to award the EUROFAB fabrication contract to COGEMA. (author)

  3. 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.

  4. 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

  5. The high moderating ratio reactor using 100% MOX reloads

    International Nuclear Information System (INIS)

    Barbrault, P.

    1994-06-01

    This report presents the concept of a High Moderating ratio Reactor, which should accept 100% MOX reloads. This reactor aims to be the plutonium version of the European Pressurized Reactor (EPR), which is developed jointly by French and German companies. A moderating ration of 2.5 (instead of the standard value of 2.0) is obtained by replacing several fuel rods by water holes. The core would contain 241 Fuel Assemblies. We present some advantages of over-moderation for plutonium fuel, a description of the core and assemblies, calculations of fuel reload schemes and Reactivity Shutdown Margins, and the behavior of the core during two occidental transients. (author). 2 refs., 9 figs., 2 tabs

  6. 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)

  7. 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.

  8. 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.

  9. 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

  10. Late effects following inhalation of mixed oxide (U,PuO2) mox aerosol in the rat

    International Nuclear Information System (INIS)

    Griffiths, N.; Van Der Meeren, A.; Fritsch, P.; Maximilien, R.

    2008-01-01

    Exposure to alpha-emitting particles is a potential long-term health risk to workers in nuclear fuel fabrication plants. Mixed Oxide (MOX: U,PuO 2 ) fuels containing low percentages of plutonium obtained from spent nuclear fuels are increasingly employed and in the case of accidental contamination by inhalation or wounds may result in the development of late-occurring pathologies such as lung cancer. However the long term risks particularly with regard to lung cancer are to date unclear. In the case of MOX the risk may indeed be different from that assigned to the individual components, plutonium and uranium. Several factors are influential (i) the dissolution of Pu depends on the physico-chemical properties, for example risk of lung cancer is increased 10 fold after Pu(NO 3 ) 2 as compared with PuO 2 . (ii) The solubility of Pu is variable whether delivered as PuO 2 or contained within MOX. (iii) The risk of cancer appears to increase with spatial homogeneity of the lung alpha dose. The objective of this study was to investigate the long term effects in rat lungs following MOX aerosol inhalation of similar particle size containing 2.5 or 7.1% Pu. Conscious rats were exposed to MOX aerosols using a 'nose-only' system and kept for their entire life (2-3 years). Different Initial Lung Deposits (ILDs) were obtained using different concentrations of the MOX suspension. Lung total alpha activity was determined in vivo at intervals over the study period by external counting as well as at autopsy in order to estimate the total lung dose. Anatomo-pathological and immunohistochemical analyses were performed on fixed lung tissue after euthanasia. The frequencies of lung pathologies and tumours were determined on lung sections at several different levels. In addition, autoradiography of lung sections was performed in order to assess the spatial localisation of a activity. Inhalation of MOX at ILD ranging from 1-20 kBq resulted in lung pathologies (90% of exposed rats

  11. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W.

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program

  12. Concept of innovative water reactor for flexible fuel cycle (FLWR)

    International Nuclear Information System (INIS)

    Iwamura, T.; Uchikawa, S.; Okubo, T.; Kugo, T.; Akie, H.; Nakatsuka, T.

    2005-01-01

    In order to ensure sustainable energy supply in the future based on the matured Light Water Reactor (LWR) and coming LWR-Mixed Oxide (MOX) technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The second part represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the fuel cycle circumstances during the reactor operation period around 60 years. At present, since the fuel cycle for the plutonium multiple recycling with MOX fuel reprocessing has not been realized yet, reprocessed plutonium from the LWR spent fuel is to be utilized in LWR-MOX. After this stage, the first part of FLWR, i.e. the high conversion type, can be introduced as a replacement of LWR or LWR-MOX. Since the plutonium inventory of FLWR is much larger, the number of the reactor with MOX fuel will be significantly reduced compared to the LWR-MOX utilization. The size of the fuel assembly for the first part is the same as in the RMWR concept, i.e. the hexagonal fuel assembly with the inner face-to-face distance of about 200 mm. Fuel rods are arranged in the triangular lattice with a relatively wide gap size around 3 mm between rods, and the effective MOX length is less than 1.5 m without using the blanket. When

  13. Polyvalent fuel treatment facility (TCP): shearing and dissolution of used fuel at La Hague facility

    Energy Technology Data Exchange (ETDEWEB)

    Brueziere, J.; Tribout-Maurizi, A.; Durand, L.; Bertrand, N. [Recycling Business Unit, AREVA, 1 place de la coupole, 92084 Paris La defense Cedex (France)

    2013-07-01

    Although many used nuclear fuel types have already been recycled, recycling plants are generally optimized for Light Water Reactor (LWR) UO{sub x} fuel. Benefits of used fuel recycling are consequently restricted to those fuels, with only limited capacity for the others like LWR MOX, Fast Reactor (FR) MOX or Research and Test Reactor (RTR) fuel. In order to recycle diverse fuel types, an innovative and polyvalent shearing and dissolving cell is planned to be put in operation in about 10 years at AREVA's La Hague recycling plant. This installation, called TCP (French acronym for polyvalent fuel treatment) will benefit from AREVA's industrial feedback, while taking part in the next steps towards a fast reactor fuel cycle development using innovative treatment solutions. Feasibility studies and R/Development trials on dissolution and shearing are currently ongoing. This new installation will allow AREVA to propose new services to its customers, in particular in term of MOX fuel, Research Test Reactors fuel and Fast Reactor fuel treatment. (authors)

  14. Thermal conductivity model of vibro-packed fuel

    International Nuclear Information System (INIS)

    Yeon Soo, Kim

    2001-01-01

    In an effort to dispose of excess weapons grade plutonium accumulated in the cold war era in the United States and the Russian Federation, one method currently under investigation is the conversion of the plutonium into mixed oxide (MOX) reactor fuel for LWRs and fast reactors in the Russian Federation. A fuel option already partly developed at the Research Institute of Atomic Reactors (RIAR) in Dimitrovgrad is that of vibro-packed MOX. Fuel rod fabrication using powder vibro-packing is attractive because it includes neither a process too complex to operate in glove boxes (or remotely), nor a waste-producing step necessary for the conventional pellet rod fabrication. However, because of its loose bonding between fuel particles at the beginning of life, vibro-packed MOX fuel has a somewhat less effective thermal conductivity than fully sintered pellet fuel, and undergoes more restructuring. Helium would also likely be pressurized in vibro-packed MOX fuel rods for LWRs to enhance initial fuel thermal conductivity. The combination of these two factors complicates development of an accurate thermal conductivity model. But clearly in order to predict fuel thermomechanical responses during irradiation of vibro-packed MOX fuel, fuel thermal conductivity must be known. The Vibropac fuel of interest in this study refers the fuel that is compacted with irregular fragments of mixed oxide fuel. In this paper, the thermal-conductivity models in the literature that dealt with relatively similar situations to the present case are examined. Then, the best model is selected based on accuracy of prediction and applicability. Then, the selected model is expanded to fit the various situations of interest. (author)

  15. 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

  16. Criticality Calculations of Fresh LEU and MOX Assemblies for Transport and Storage at the Balakovo Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Goluoglu, S.

    2001-01-11

    Transportation of low-enriched uranium (LEU) and mixed-oxide (MOX) assemblies to and within the VVER-1000-type Balakovo Nuclear Power Plant is investigated. Effective multiplication factors for fresh fuel assemblies on the railroad platform, fresh fuel assemblies in the fuel transportation vehicle, and fresh fuel assemblies in the spent fuel storage pool are calculated. If there is no absorber between the units, the configurations with all MOX assemblies result in higher effective multiplication factors than the configurations with all LEU assemblies when the system is dry. When the system is flooded, the configurations with all LEU assemblies result in higher effective multiplication factors. For normal operating conditions, effective multiplication factors for all configurations are below the presumed upper subcritical limit of 0.95. For an accident condition of a fully loaded fuel transportation vehicle that is flooded with low-density water (possibly from a fire suppression system), the presumed upper subcritical limit is exceeded by configurations containing LEU assemblies.

  17. Isotopic modelling using the ENIGMA-B fuel performance code

    International Nuclear Information System (INIS)

    Rossiter, G.D.; Cook, P.M.A.; Weston, R.

    2001-01-01

    A number of experimental programmes by BNFL and other MOX fabricators have now shown that the in-pile performance of MOX fuel is generally similar to that of conventional UO 2 fuel. Models based on UO 2 fuel experience form a good basis for a description of MOX fuel behaviour. However, an area where the performance of MOX fuel is sufficiently different from that of UO 2 to warrant model changes is in the radial power and burnup profile. The differences in radial power and burnup profile arise from the presence of significant concentrations of plutonium in MOX fuel, at beginning of life, and their subsequent evolution with burnup. Amongst other effects, plutonium has a greater neutron absorption cross-section than uranium. This paper focuses on the development of a new model for the radial power and burnup profile within a UO 2 or MOX fuel rod, in which the underlying fissile isotope concentration distributions are tracked during irradiation. The new model has been incorporated into the ENIGMA-B fuel performance code and has been extended to track the isotopic concentrations of the fission gases, xenon and krypton. The calculated distributions have been validated against results from rod puncture measurements and electron probe micro-analysis (EPMA) linescans, performed during the M501 post irradiation examination (PIE) programme. The predicted gas inventory of the fuel/clad gap is compared with the isotopic composition measured during rod puncture and the measured radial distributions of burnup (from neodymium measurements) and plutonium in the fuel are compared with the calculated distributions. It is shown that there is good agreement between the code predictions and the measurements. (author)

  18. 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)

  19. Mixed U/Pu oxide fabrication facility for gel-sphere-pac fuel

    International Nuclear Information System (INIS)

    1978-09-01

    This paper describes a conceptual plant which uses the gel-sphere-pac process to fabricate mixed oxide (MOX) fuel and covers (1) fabrication of co-processed MOX fuel and (2) fabrication of co-processed spiked MOX fuel, using 60 Co. The report describes: the fuel fabrication process and plant layout, including scrap and waste processing; and maintenance safety and ventilation measures. A description of the conversion of U and Pu nitrate using a gel sphere process is given in Appendix A

  20. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  1. MOX manufacturing perspectives in a fast growing future and the MELOX plant

    International Nuclear Information System (INIS)

    Bekiarian, A.; Le Bastard, G.

    1991-01-01

    The potential MOX fuel market will grow regularly in the nineties. In view of satisfying the needs of the market, mixed-oxide fuel manufacturers have a strong incentive to increase the capacity of existing facilities and to build new ones. The Belgonucleaire plant at Dessel has been in operation since 1973. It has been backfitted up to a capacity of 35 t/y of LWR fuel which is now fully available. To satisfy the need of MOX fuel it was equally decided to adapt facilities in Cadarache where a production line, with a capacity of 15 t/y, is now delivering its production. But planned production up to the end of the century implies further increases in manufacturing capacities : MELOX, a plant for 120 t/y is under construction on the COGEMA site of Marcoule as well as a further expansion of Belgonucleaire plant at Dessel (P1) is studied to reach 70 t/y on this site. Similar developments are also planned by SIEMENS for a new manufacturing capability at Hanau (Germany). MELOX as well as all the new facilities have to get high levels of safety concerning environment and personnel. This leads to largely automated operations, and a particular care for waste treatment. (author)

  2. 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)

  3. Conceptual study of the future nuclear fuel cycle system for the extended LWR age

    International Nuclear Information System (INIS)

    Fujine, Sachio; Takano, Hideki; Sato, Osamu; Tone, Tatsuzo; Yamada, Takashi; Kurosawa, Katsutoshi.

    1993-08-01

    A large scale integrated fuel cycle facility (IFCF) is assumed for the future nuclear fuel cycle in the extended LWR age. Spent MOX fuels are reprocessed mixed with UOX in a centralized reprocessing plant. The reprocessing plant separates long-lived nuclides as well as Pu. Nitric acid solutions of those products are fed directly to MOX fabrication process which is incorporated with reprocessing. MOX pellets are made by sphere-cal process. Two process concepts are made as advanced reprocessing incorporated with partitioning (ARP) which has the function of long-lived nuclides recovery. One is a simplified Purex combined with partitioning. Extractable long-lived nuclides, 237 Np and 99 Tc, are assumed to be recovered in main flow stream of the improved Purex process. The other process concept is made aiming at recovering all TRU nuclides in reprocessing to meet with TRU recycle requirement in the long future. A concept of the future fuel cycle system is made by combining integrated fuel cycle facility and very high burnup LWRs (VHBR). The reactor concept of VHBRs has been proposed to improve Pu recycle economy in the future. Highly enriched MOX fuel are loaded in the full core of reactor in order to increase reactivity for the burnup. Fuel cycle indices such as Pu isotopic composition change, spent fuel integration, nuclide transmutation effect are estimated by simulating the Pu recycling in the system of VHBR and ARP. It is concluded that Pu enrichment of MOX fuel can be kept less than 20 % through multi-recycle. Reprocessing MOX fuels with UOX shows a favorable effect for keeping Pu reactivity high enough for VHBR. Integration of spent MOX fuel can be reduced by Pu recycle. Transmutation of Np is feasible by containing Np into MOX fuel. (author)

  4. 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

  5. Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Chang, G.S.; Ryskamp, J.M.; Terry, W.K.; Ambrosek, R.G.; Palmer, A.J.; Roesener, R.A.

    1996-09-01

    The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

  6. Environmental assessment for the manufacture and shipment of nuclear reactor fuel from the United States to Canada

    International Nuclear Information System (INIS)

    Rangel, R.C.

    1999-01-01

    The US Department of Energy (DOE) has declared 41.9 tons (38 metric tons) of weapons-usable plutonium surplus to the United States' defense needs. A DOE Programmatic Environmental Impact Statement analyzed strategies for plutonium storage and dispositioning. In one alternative, plutonium as a mixed oxide (MOX) fuel would be irradiated (burned) in a reengineered heavy-water-moderated reactor, such as the Canadian CANDU design. In an Environmental Assessment (EA), DOE proposes to fabricate and transport to Canada a limited amount of MOX fuel as part of the Parallex (parallel experiment) Project. MOX fuel from the US and Russia would be used by Canada to conduct performance tests at Chalk River Laboratories. MOX fuel would be fabricated at Los Alamos National Laboratory and transported in approved container(s) to a Canadian port(s) of entry on one to three approved routes. The EA analyzes the environmental and human health effects from MOX fuel fabrication and transportation. Under the Proposed Action, MOX fuel fabrication would not result in adverse effects to the involved workers or public. Analysis showed that the shipment(s) of MOX fuel would not adversely affect the public, truck crew, and environment along the transportation routes

  7. 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

  8. Uncertainty Analysis of Light Water Reactor Fuel Lattices

    Directory of Open Access Journals (Sweden)

    C. Arenas

    2013-01-01

    Full Text Available The study explored the calculation of uncertainty based on available cross-section covariance data and computational tool on fuel lattice levels, which included pin cell and the fuel assembly models. Uncertainty variations due to temperatures changes and different fuel compositions are the main focus of this analysis. Selected assemblies and unit pin cells were analyzed according to the OECD LWR UAM benchmark specifications. Criticality and uncertainty analysis were performed using TSUNAMI-2D sequence in SCALE 6.1. It was found that uncertainties increase with increasing temperature, while kinf decreases. This increase in the uncertainty is due to the increase in sensitivity of the largest contributing reaction of uncertainty, namely, the neutron capture reaction 238U(n, γ due to the Doppler broadening. In addition, three types (UOX, MOX, and UOX-Gd2O3 of fuel material compositions were analyzed. A remarkable increase in uncertainty in kinf was observed for the case of MOX fuel. The increase in uncertainty of kinf in MOX fuel was nearly twice the corresponding value in UOX fuel. The neutron-nuclide reaction of 238U, mainly inelastic scattering (n, n′, contributed the most to the uncertainties in the MOX fuel, shifting the neutron spectrum to higher energy compared to the UOX fuel.

  9. Determination of chloride in MOX samples using chloride ion selective electrode

    Energy Technology Data Exchange (ETDEWEB)

    Govindan, R; Das, D K; Mallik, G K; Sumathi, A; Patil, Sangeeta; Raul, Seema; Bhargava, V K; Kamath, H S [Bhabha Atomic Research Centre, Tarapur (India). Advanced Fuel Fabrication Facility

    1997-09-01

    The chloride present in the MOX fuel is separated from the matrix by pyrohydrolysis at a temperature of 950 {+-} 50 degC and is then analyzed by chloride ion selective electrode (Cl-ISE). The range covered is 0.4-4 ppm with a precision of better than {+-}5% R.S.D. (author). 4 refs., 1 tab.

  10. Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2005-07-01

    At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

  11. Recycling of plutonium and uranium in water reactor fuel. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-05-01

    The Technical Committee Meeting on Recycling of Plutonium and Uranium in Water Reactor Fuel was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its aim was to obtain an overall picture of MOX fabrication capacity and technology, actual performance of this kind of fuel, and ways explored to dispose of the weapons grade plutonium. The subject of this meeting had been reviewed by the International Atomic Energy Agency every 5 to 6 years and for the first time the problem of weapons grade plutonium disposal was included. The papers presented provide a summary of experience on MOX fuel and ongoing research in this field in the participating countries. The meeting was hosted by British Nuclear Fuels plc, at Newby Bridge, United Kingdom, from 3 to 7 July 1995. Fifty-six participants from twelve countries or international organizations took part. Refs, figs, tabs

  12. Progress of full MOX core design in ABWR

    International Nuclear Information System (INIS)

    Izutsu, S.; Sasagawa, M.; Aoyama, M.; Maruyama, H.; Suzuki, T.

    2000-01-01

    Full MOX ABWR core design has been made, based on the MOX design concept of 8x8 bundle configuration with a large central water rod, 40 GWd/t maximum bundle exposure, and the compatibility with 9x9 high-burnup UO 2 bundles. Core performance on shutdown margin and thermal margin of the MOX-loaded core is similar to that of UO 2 cores for the range from full UO 2 core to full MOX core. Safety analyses based on its safety parameters and MOX property have shown its conformity to the design criteria in Japan. In order to confirm the applicability of the nuclear design method to full MOX cores, Tank-type Critical Assembly (TCA) experiment data have been analyzed on criticality, power distribution and β eff /l measurements. (author)

  13. Late effects following inhalation of mixed oxide (U,PuO{sub 2}) mox aerosol in the rat; Effets tardifs de l'inhalation d'aerosols de Mox 2,5% ou 7,1% Pu chez le rat

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, N.; Van Der Meeren, A.; Fritsch, P.; Maximilien, R

    2008-07-01

    Exposure to alpha-emitting particles is a potential long-term health risk to workers in nuclear fuel fabrication plants. Mixed Oxide (MOX: U,PuO{sub 2}) fuels containing low percentages of plutonium obtained from spent nuclear fuels are increasingly employed and in the case of accidental contamination by inhalation or wounds may result in the development of late-occurring pathologies such as lung cancer. However the long term risks particularly with regard to lung cancer are to date unclear. In the case of MOX the risk may indeed be different from that assigned to the individual components, plutonium and uranium. Several factors are influential (i) the dissolution of Pu depends on the physico-chemical properties, for example risk of lung cancer is increased 10 fold after Pu(NO{sub 3}){sub 2} as compared with PuO{sub 2}. (ii) The solubility of Pu is variable whether delivered as PuO{sub 2} or contained within MOX. (iii) The risk of cancer appears to increase with spatial homogeneity of the lung alpha dose. The objective of this study was to investigate the long term effects in rat lungs following MOX aerosol inhalation of similar particle size containing 2.5 or 7.1% Pu. Conscious rats were exposed to MOX aerosols using a 'nose-only' system and kept for their entire life (2-3 years). Different Initial Lung Deposits (ILDs) were obtained using different concentrations of the MOX suspension. Lung total alpha activity was determined in vivo at intervals over the study period by external counting as well as at autopsy in order to estimate the total lung dose. Anatomo-pathological and immunohistochemical analyses were performed on fixed lung tissue after euthanasia. The frequencies of lung pathologies and tumours were determined on lung sections at several different levels. In addition, autoradiography of lung sections was performed in order to assess the spatial localisation of a activity. Inhalation of MOX at ILD ranging from 1-20 kBq resulted in lung

  14. International collaborations about fuel studies for reactor recycling of military quality plutonium

    International Nuclear Information System (INIS)

    Bernard, H.; Chaudat, J.P.

    1997-01-01

    In November 1992, an agreement was signed between the French and Russian governments to use in Russia and for pacific purposes the plutonium recovered from the Russian nuclear weapons dismantling. This plutonium will be transformed into mixed oxide fuels (MOX) for nuclear power production. The French Direction of Military Applications (DAM) of the CEA is the operator of the French-Russian AIDA program. The CEA Direction of Fuel Cycle (DCC) and Direction of Nuclear Reactors (DRN) are involved in the transformation of metallic plutonium into sinterable oxide powder for MOX fuel manufacturing. The Russian TOMOX (Treatment of MOX powder Metallic Objects) and DEMOX (MOX Demonstration) plants will produce the MOX fuel assemblies for the 4 VVER 1000 reactors of Balakovo and the fast BN 600 reactor. The second part of the program will involve the German Siemens and GRS companies for the safety studies of the reactors and fuel cycle plants. The paper gives also a brief analysis of the US policy concerning the military plutonium recycling. (J.S.)

  15. Fuel qualification issues and strategies for reactor-based surplus plutonium disposition

    International Nuclear Information System (INIS)

    Cowell, B.S.; Copeland, G.L.; Moses, D.L.

    1997-08-01

    The Department of Energy (DOE) has proposed irradiation of mixed-oxide (MOX) fuel in existing commercial reactors as a disposition method for surplus plutonium from the weapons program. The burning of MOX fuel in reactors is supported by an extensive technology base; however, the infrastructure required to implement reactor-based plutonium disposition does not exist domestically. This report identifies and examines the actions required to qualify and license weapons-grade (WG) plutonium-based MOX fuels for use in domestic commercial light-water reactors (LWRs)

  16. Uranium-plutonium fuel for fast reactors

    International Nuclear Information System (INIS)

    Antipov, S.A.; Astafiev, V.A.; Clouchenkov, A.E.; Gustchin, K.I.; Menshikova, T.S.

    1996-01-01

    Technology was established for fabrication of MOX fuel pellets from co-precipitated and mechanically blended mixed oxides. Both processes ensure the homogeneous structure of pellets readily dissolvable in nitric acid upon reprocessing. In order to increase the plutonium charge in a reactor-burner a process was tested for producing MOX fuel with higher content of plutonium and an inert diluent. It was shown that it is feasible to produce fuel having homogeneous structure and the content of plutonium up to 45% mass

  17. Evaluation of remaining behavior of halogen on the fabrication of MOX pellet containing Am

    International Nuclear Information System (INIS)

    Ozaki, Yoko; Osaka, Masahiko; Obayashi, Hiroshi; Tanaka, Kenya

    2004-11-01

    It is important to limit the content of halogen elements, namely fluorine and chlorine that are sources of making cladding material corrode, in nuclear fuel from the viewpoint of quality assurance. The halogen content should be more carefully limited in the MOX fuel containing Americium (Am-MOX), which is fabricated in the Alpha-Gamma Facility (AGF) for irradiation testing to be conducted in the experimental fast reactor JOYO, because fluorine may remain in the sintered pellets owing to a formation of AmF 3 known to have a low vapor pressure and may exceeds the limit of 25 ppm. In this study, a series of experimental determination of halogen element in Am-MOX were performed by a combination method of pyrolysis and ion-chromatography for the purpose of an evaluation of behavior of remaining halogen through the sintering process. Oxygen potential, temperature and time were changed as experimental parameters and their effects on the remaining behavior of halogen were examined. It was confirmed that good pellets, which contained small amount of halogen, could be obtained by the sintering for 3 hour at 1700degC in the oxygen potential range from -520 to -390 kJ/mol. In order to analysis of fluorine chemical form in green pellet, thermal analysis was performed. AmF 3 and PuF 3 have been confirmed to remain in the green pellet. (author)

  18. Licensing issues associated with the use of mixed-oxide fuel in U.S. commercial nuclear reactors

    International Nuclear Information System (INIS)

    Williams, D.L. Jr.

    1997-04-01

    On January 14, 1997, the Department of Energy, as part of its Record of Decision on the storage and disposition of surplus nuclear weapons materials, committed to pursue the use of excess weapons-usable plutonium in the fabrication of mixed-oxide (MOX) fuel for consumption in existing commercial nuclear power plants. Domestic use of MOX fuel has been deferred since the late 1970s, principally due to nuclear proliferation concerns. This report documents a review of past and present literature (i.e., correspondence, reports, etc.) on the domestic use of MOX fuel and provides discussion on the technical and regulatory issues that must be addressed by DOE (and the utility/consortia selected by DOE to effect the MOX fuel consumption strategy) in obtaining approval from the Nuclear Regulatory Commission to use MOX fuel in one or a group of existing commercial nuclear power plants

  19. 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)

  20. 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)

  1. 43 CFR 2.47 - Records subject to Privacy Act.

    Science.gov (United States)

    2010-10-01

    ... 43 Public Lands: Interior 1 2010-10-01 2010-10-01 false Records subject to Privacy Act. 2.47 Section 2.47 Public Lands: Interior Office of the Secretary of the Interior RECORDS AND TESTIMONY; FREEDOM OF INFORMATION ACT Privacy Act § 2.47 Records subject to Privacy Act. The Privacy Act applies to all...

  2. VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, RJ

    2001-02-02

    The Task Force on Reactor-Based Plutonium Disposition, now an Expert Group, was set up through the Organization for Economic Cooperation and Development/Nuclear Energy Agency to facilitate technical assessments of burning weapons-grade plutonium mixed-oxide (MOX) fuel in U.S. pressurized-water reactors and Russian VVER nuclear reactors. More than ten countries participated to advance the work of the Task Force in a major initiative, which was a blind benchmark study to compare code benchmark calculations against experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At the Oak Ridge National Laboratory, the HELIOS-1.4 code was used to perform a comprehensive study of pin-cell and core calculations for the VENUS-2 benchmark.

  3. Nuclear fuel activities in Canada

    Energy Technology Data Exchange (ETDEWEB)

    Cox, D S [Fuel Development Branch, Chalk River Labs., AECL (Canada)

    1997-12-01

    Nuclear fuel activities in Canada are considered in the presentation on the following directions: Canadian utility fuel performance; CANDU owner`s group fuel programs; AECL advanced fuel program (high burnup fuel behaviour and development); Pu dispositioning (MOX) activities. 1 tab.

  4. Microwave based oxidation process for recycling the off-specification (U,Pu)O{sub 2} fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Singh, G., E-mail: gitendars@barctara.gov.in [Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, 401 502 (India); Khot, P.M. [Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, 401 502 (India); Kumar, Pradeep [Integrated Fuel Fabrication Facility (IFFF), Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Bhatt, R.B.; Behere, P.G.; Afzal, Mohd [Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, 401 502 (India)

    2017-02-15

    This paper reports development of a process named MicroWave Direct Oxidation (MWDO) for recycling the off-specification (U,Pu)O{sub 2} mixed oxide (MOX) fuel pellets generated during fabrication of typical fast reactor fuels. MWDO is a two-stage, single-cycle process based on oxidative pulverisation of pellets using 2450 MHz microwave. The powder sinterability was evaluated by bulk density and BET specific surface area. The oxidised powders were analyzed for phases using XRD and stoichiometry by thermogravimetry. The sinterability was significantly enhanced by carrying out oxidation in higher oxygen partial pressure and by subjecting MOX to multiple micronisation-oxidation cycles. After three cycles, the recycled powder from (U,28%Pu)O{sub 2} resulted surface area >3 m{sup 2}/g and 100% re-used for MOX fabrication. The flow sheet was developed for maximum utilization of recycled powder describable by a parameter called Scrap Recycling Ratio (SRR). The process demonstrates smaller processing cycle, better powder properties and higher oxidative pulverisation over conventional method. - Highlights: • A process for recycling the off-specification (U,Pu)O{sub 2} sintered fuel pellets of fast reactors was demonstrated. • The method is a two-stage, single cycle process based on oxidative pulverization of MOX pellets using 2450 MHz microwave. • The process demonstrated utilization of recycled powder with SRR of 1.

  5. Status of core nuclear design technology for future fuel

    International Nuclear Information System (INIS)

    Joo, Hyung Kook; Jung, Hyung Guk; Noh, Jae Man; Kim, Yeong Il; Kim, Taek Kyum; Gil, Choong Sup; Kim, Jung Do; Kim, Young Jin; Sohn, Dong Seong

    1997-01-01

    The effective utilization of nuclear resource is more important factor to be considered in the design of next generation PWR in addition to the epochal consideration on economics and safety. Assuming that MOX fuel can be considered as one of the future fuel corresponding to the above request, the establishment of basic technology for the MOX core design has been performed : : the specification of the technical problem through the preliminary core design and nuclear characteristic analysis of MOX, the development and verification of the neutron library for lattice code, and the acquisition of data to be used for verification of lattice and core analysis codes. The following further studies will be done in future: detailed verification of library E63LIB/A, development of the spectral history effect treatment module, extension of decay chain, development of new homogenization for the MOX fuel assembly. (author). 6 refs., 7 tabs., 2 figs

  6. 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)

  7. Imminent: Irradiation Testing of (Th,Pu)O{sub 2} Fuel - 13560

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Julian F. [Thor Energy AS, Sommerrogaten 13-15, Oslo 0255 (Norway); Franceschini, Fausto [Westinghouse Electric Company LLC, 1000 Cranberry Woods Drive, Cranberry Township, PA 16066 (United States)

    2013-07-01

    Commercial-prototype thorium-plutonium oxide (Th-MOX) fuel pellets have been loaded into the material test reactor in Halden, Norway. The fuel is being operated at full power - with instrumentation - in simulated LWR / PHWR conditions and its behaviour is measured 'on-line' as it operates to high burn-up. This is a vital test on the commercialization pathway for this robust new thoria-based fuel. The performance data that is collected will support a fuel modeling effort to support its safety qualification. Several different samples of Th-MOX fuel will be tested, thereby collecting information on ceramic behaviours and their microstructure dependency. The fuel-cycle reasoning underpinning the test campaign is that commercial Th- MOX fuels are an achievable intermediate / near-term SNF management strategy that integrates well with a fast reactor future. (authors)

  8. Some alternatives to the mixed oxide fuel cycle

    International Nuclear Information System (INIS)

    Deonigi, D.E.; Eschbach, E.A.; Goldsmith, S.; Pankaskie, P.J.; Rohrmann, C.A.; Widrig, R.D.

    1977-02-01

    While on initial examination each of the six fuel cycle concepts (tandem cycle, extended burnup, fuel rejuvenation, coprocessing, partial reprocessing, and thorium) described in the report may have some potential for improving safeguards, none of the six appears to have any other major or compelling advantages over the mixed oxide (MOX) fuel cycle. Compared to the MOX cycle, all but coprocessing appear to have major disadvantages, including severe cost penalties. Three of the concepts-tandem, extended burnup, and rejuvenation--share the basic problems of the throwaway cycle (GESMO Alternative 6): without reprocessing, high-level waste volumes and costs are substantially increased, and overall uranium utilization decreases for three reasons. First, the parasitic fission products left in the fuel absorb neutrons in later irradiation steps reducing the overall neutronic efficiencies of these cycles. Second, discarded fuel still has sufficient fissile values to warrant recycle. Third, perhaps most important, the plutonium needed for breeder start-up will not be available; without the breeder, uranium utilization would drop by about a factor of sixty. Two of the concepts--coprocessing and partial reprocessing--involve variations of the basic MOX fuel cycle's chemical reprocessing step to make plutonium diversion potentially more difficult. These concepts could be used with the MOX fuel cycle or in conjunction with the tandem, extended burnup and rejuvenation concepts to eliminate some of the problems with those cycles. But in so doing, the basic impetus for those cycles--elimination of reprocessing for safeguards purposes--no longer exists. Of all the concepts considered, only coprocessing--and particularly the ''master blend'' version--appears to have sufficient promise to warrant a more detailed study. The master blend concept could possibly make plutonium diversion more difficult with minimal impact on the reprocessing and MOX fuel fabrication operations

  9. Experience on Russian military origin plutonium conversion into fast reactor nuclear fuel

    International Nuclear Information System (INIS)

    Grachev, A.F.; Skiba, O.V.; Bychkov, A.V.; Mayorshin, A.A.; Kisly, V.A.; Bobrov, D.A.; Osipenko, A.G.; Babikov, L.G.; Mishinev, V.B.

    2001-01-01

    According to the Concept of Russian Minatom on military plutonium excess utilization, the State Scientific Center of Russian Federation ''Research Institute of Atomic Reactors'' (Dimitrovgrad) has begun study on possibility of technological processing of the metal military plutonium into MOX fuel. The Program and the stages of its realization are submitted in the paper. During 1998-2000 the first stage of the Program was fulfilled and 50 kg of military origin metallic plutonium was converted to MOX fuel for the BOR-60 and BN-600 reactor. The plutonium conversion into MOX fuel is carried out under the original technology developed by SSC RIAR. It includes pyro-electrochemical process for production of fuel on the domestic equipment with the subsequent fuel pins manufacturing for the fast reactors by the vibro-packing method. The produced MOX fuel is purified from alloy additives (Ga) and corresponds to the vibro-packed fuel standard for fast reactors. The fuel pins manufacturing for BOR-60 and BN-600 reactors are carried out by the vibro-packing method on a standard procedure, which is used in SSC RIAR more than 20 years. (author)

  10. Light water reactor mixed-oxide fuel irradiation experiment

    International Nuclear Information System (INIS)

    Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

    1998-01-01

    The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

  11. Neutronic feasibility of PWR core with mixed oxide fuels in the Republic of Korea

    International Nuclear Information System (INIS)

    Kim, Y.J.; Joo, H.K.; Jung, H.G.; Sohn, D.S.

    1997-01-01

    Neutronic feasibility of a PWR core with mixed oxide (MOX) fuels has been investigated as part of the feasibility study for recycling spent fuels in Korea. A typical 3-loop PWR with 900 MWe capacity is selected as reference plant to develop equilibrium core designs with low-leakage fuel management scheme, while incorporating various MOX loading. The fuel management analyses and limited safety analyses show that, safely stated, MOX recycling with 1/3 reload fraction can be accommodated for both annual and 18 month fuel cycle schemes in Korean PWRs, without major design modifications on the reactor systems. (author). 12 refs, 4 figs, 3 tabs

  12. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    International Nuclear Information System (INIS)

    Johnson, E.R.; Best, R.E.

    2009-01-01

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the 'UREX+3c fuel cycle' and the 'Alternative Fuel Cycle' (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount of the

  13. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  14. Evaluation of thermal physical properties for fast reactor fuels. Melting point and thermal conductivities

    International Nuclear Information System (INIS)

    Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki; Abe, Tomoyuki; Uno, Hiroki; Ogasawara, Masahiro; Tamura, Tetsuya; Sugata, Hirotada; Sunaoshi, Takeo; Shibata, Kazuya

    2006-10-01

    Japan Atomic Energy Agency has developed a fast breeder reactor (FBR), and plutonium and uranium mixed oxide (MOX) having low density and 20-30%Pu content has used as a fuel of the FBR, Monju. In plutonium, Americium has been accumulated during long-term storage, and Am content will be increasing up to 2-3% in the MOX. It is essential to evaluate the influence of Am content on physical properties of MOX on the development of FBR in the future. In this study melting points and thermal conductivities which are important data on the fuel design were measured systematically in wide range of composition, and the effects of Am accumulated were evaluated. The solidus temperatures of MOX were measured as a function of Pu content, oxygen to metal ratio (O/M) and Am content using thermal arrest technique. The sample was sealed in a tungsten capsule in vacuum for measuring solidus temperature. In the measurements of MOX with Pu content of more than 30%, a rhenium inner capsule was used to prevent the reaction between MOX and tungsten. In the results, it was confirmed that the melting points of MOX decrease with as an increase of Pu content and increase slightly with a decrease of O/M ratio. The effect of Am content on the fuel design was negligible small in the range of Am content up to 3%. Thermal conductivities of MOX were evaluated from thermal diffusivity measured by laser flash method and heat capacity calculated by Neumann- Kopp's law. The thermal conductivity of MOX decreased slightly in the temperature of less than 1173K with increasing Am content. The effect of Am accumulated in long-term storage fuel was evaluated from melting points and thermal conductivities measured in this study. It is concluded that the increase of Am in the fuel barely affect the fuel design in the range of less than 3%Am content. (author)

  15. 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)

  16. Links among available integral benchmarks and differential date evaluations, computational biases and uncertainties, and nuclear criticality safety biases on potential MOX production throughput

    International Nuclear Information System (INIS)

    Goluoglu, S.; Hopper, C.M.

    2004-01-01

    Through the use of Oak Ridge National Laboratory's recently developed and applied sensitivity and uncertainty computational analysis techniques, this paper presents the relevance and importance of available and needed integral benchmarks and differential data evaluations impacting potential MOX production throughput determinations relative to low-moderated MOX fuel blending operations. The relevance and importance in the availability of or need for critical experiment benchmarks and data evaluations are presented in terms of computational biases as influenced by computational and experimental sensitivities and uncertainties relative to selected MOX production powder blending processes. Recent developments for estimating the safe margins of subcriticality for assuring nuclear criticality safety for process approval are presented. In addition, the impact of the safe margins (due to computational biases and uncertainties) on potential MOX production throughput will also be presented. (author)

  17. Safety characteristics of mid-sized MOX fueled liquid metal reactor core of high converter type in the initiating phase of unprotected loss of flow accident. Effect of low specific fuel power density on ULOF behavior brought by employment of large diameter fuel pins

    International Nuclear Information System (INIS)

    Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime

    2003-07-01

    Safety characteristics in core disruptive accidents (CDAs) of mid-sized MOX fueled liquid metal reactor core of high converter type have been examined by using the CDA initiating phase analysis code SAS4A. The design concept of high converter type reactor core has been studied as one of options in the category of sodium-cooled reactor in Phase II of Feasibility Study on Commercialized Fast Reactor Cycle System. An unprotected loss-of-flow accident (ULOF) has been selected as a representative CDA initiator for this study. A core concept of high converter type, which employed a large diameter fuel pin of 11.1 mm with 1.2 m core height to get a large fuel volume fraction in the core to achieve high internal conversion ratio was proposed in JFY2001. Each fuel subassembly of the core (abbreviated here as UPL120)was provided with an upper sodium plenum directly above the core to reduce the sodium void reactivity worth. Because of the large fuel pin diameter, average specific fuel power density (31 kW/kg-MOX) of UPL120 is about one half of those of conventional large MOX cores. The reactivity worth of sodium voiding is 6$ in the whole core, and -1$ in the all upper plenums. Initiating phase of ULOF accident in UPL120 under the conditions of nominal design and best estimate analysis resulted in a slightly super-prompt critical power burst. The causes of the super-prompt criticality have been identified twofold: (a) the low specific fuel power density of core reduced the effectiveness of prompt negative reactivity feedback of Doppler and axial fuel expansion effects upon increase in reactor power, and (b) the longer core height compared with conventional 1m cores brought, together with the lower specific power density, a remarkable delay in insertion of negative fuel dispersion reactivity after the onset of fuel disruption in sodium voided subassembly due to the lower linear heat rating in the top portion of the core. During the delay, burst-type fuel failures in sodium un

  18. AFCI : Co-extraction impacts on LWR and fast reactor fuel cycles

    International Nuclear Information System (INIS)

    Taiwo, T. A.; Szakalay, F. J.; Kim, T. K.; Hill, R. N.; Nuclear Engineering Division

    2007-01-01

    A systematic investigation of the impact of the co-extraction COEXTM process on reactor performance has been performed. The proliferation implication of the process was also evaluated using the critical mass, radioactivity, decay heat and neutron and gamma source rates and gamma doses as indicators. The use of LWR-spent-uranium-based MOX fuel results in a higher initial plutonium content requirement in an LWR MOX core than if natural uranium based MOX fuel is used (by about 1%); the plutonium for both cases is derived from the spent LWR spent fuel. More transuranics are consequently discharged in the spent fuel of the MOX core. The presence of U-236 in the initial fuel was also found to result in higher content of Np-237 in the spent MOX fuel and less consumption of Pu-238 and Am-241 in the MOX core. The higher quantities of Np-237 (factor of 5), Pu-238 (20%) and Am-241 (14%) decrease the effective repository utilization, relative to the use of natural uranium in the PWR MOX core. Additionally, the minor actinides continue to accumulate in the fuel cycle, even if the U-Pu co-extraction products are continuously recycled in the PWR cores, and thus a solution is required for the minor actinides. The utilization of plutonium derived from LWR spent fuel versus weapons-grade plutonium for the startup core of a 1,000 MWT advanced burner fast reactor (ABR) increases the TRU content by about 4%. Differences are negligible for the equilibrium recycle core. The impact of using reactor spent uranium instead of depleted uranium was found to be relatively smaller in the fast reactor (TRU content difference less than 0.4%). The critical masses of the co-extraction products were found to be higher than that of weapons-grade plutonium and the decay heat and radiation sources of the materials (products) were also found to be generally higher than that of weapons-grade plutonium (WG-Pu) in the transuranics content range of 0.1 to 1.0 in the heavy-metal. The magnitude of the

  19. 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)

  20. Potential of thorium-based fuel cycle for PWR core to reduce plutonium and long-term toxicity

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Kim, Taek Kyum; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    The cross section libraries and calculation methods of the participants were inter-compared through the first stage benchmark calculation. The multiplication factor of unit cell benchmark are in good agreement, but there is significant discrepancies of 2.3 to 3.5 %k at BOC and at EOC between the calculated infinite multiplication factors of each participants for the assembly benchmark. Our results with HELIOS show a reasonable agreement with the others except the MTC value at EOC. To verify the potential of the thorium-based fuel to consume the plutonium and to reduce the radioactivity from the spent fuel, the conceptual core with ThO{sub 2}-PuO{sub 2} or MOX fuel were constructed. The composition and quantity of plutonium isotopes and the radioactivity level of spent fuel for conceptual cores were analyzed, and the neutronic characteristics of conceptual cores were also calculated. The nuclear characteristics for ThO{sub 2}-PuO{sub 2} thorium fueled core was similar to MOX fueled core, mainly due to the same seed fuel material, plutonium. For the capability of plutonium consumption, ThO{sub 2}-PuO{sub 2} thorium fuel can consume plutonium 2.1-2.4 times MOX fuel. The fraction of fissile plutonium in the spent ThO{sub 2}-PuO{sub 2} thorium fuel is more favorable in view of plutonium consumption and non-proliferation than MOX fuel. The radioactivity of spent ThO{sub 2}-PuO{sub 2} thorium and MOX fuel batches were calculated. Since plutonium isotopes are dominant for the long-term radioactivity, ThO{sub 2}-PuO{sub 2} thorium has almost the same level of radioactivity as in MOX fuel for a long-term perspective. (author). 22 figs., 11 tabs.

  1. Nuclear fuels

    International Nuclear Information System (INIS)

    2008-01-01

    The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

  2. 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)

  3. Pt/MOx/SiO2, Pt/MOx/TiO2, and Pt/MOx/Al2O3 Catalysts for CO Oxidation

    Directory of Open Access Journals (Sweden)

    Hongmei Qin

    2015-04-01

    Full Text Available Conventional supported Pt catalysts have often been prepared by loading Pt onto commercial supports, such as SiO2, TiO2, Al2O3, and carbon. These catalysts usually have simple metal-support (i.e., Pt-SiO2 interfaces. To tune the catalytic performance of supported Pt catalysts, it is desirable to modify the metal-support interfaces by incorporating an oxide additive into the catalyst formula. Here we prepared three series of metal oxide-modified Pt catalysts (i.e., Pt/MOx/SiO2, Pt/MOx/TiO2, and Pt/MOx/Al2O3, where M = Al, Fe, Co, Cu, Zn, Ba, La for CO oxidation. Among them, Pt/CoOx/SiO2, Pt/CoOx/TiO2, and Pt/CoOx/Al2O3 showed the highest catalytic activities. Relevant samples were characterized by N2 adsorption-desorption, X-ray diffraction (XRD, transmission electron microscopy (TEM, H2 temperature-programmed reduction (H2-TPR, X-ray photoelectron spectroscopy (XPS, CO temperature-programmed desorption (CO-TPD, O2 temperature-programmed desorption (O2-TPD, and CO2 temperature-programmed desorption (CO2-TPD.

  4. Technical design considerations in the provision of a commercial MOX plant

    International Nuclear Information System (INIS)

    Elliott, M.F.

    1997-01-01

    The Sellafield MOX Plant (SMP) has a design production target of 120 t/year Heavy Metal of mixed uranium dioxide and plutonium dioxided (MOX) fuel. It will have the capability to produce fuel with fissile enrichments up to 10%. The feed materials are those arising from reprocessing operations on the Sellafield site, although the plant also has the capability to receive and process plutonium from overseas reprocessing plants. The ability to produce 10% enriched fuels, together with the requirement to use high burn-up feed has posed a number of design challenges to prevent excessive powder temperatures within the plant. As no stimulants are available to represent the heat generating nature of plutonium powders, it is difficult to prove equipment design by experiment. Extensive use has therefore been made of finite element analysis techniques. The requirement to process material of low burn-up (i.e. high fissile enrichment) has also impacted on equipment design in order to ensure that criticality limits are not exceeded. This has been achieved where possible by 'safe by geometry' design and, where appropriate, by high integrity protection systems. SMP has been designed with a high plant availability but at minimum cost. The requirement to minimize cost has meant that high availability must be obtained with the minimum of equipment. This had led to major challenges for equipment designers in terms of both the reliability and also the maintainability of equipment. Extensive use has been made of theoretical modelling techniques which have given confidence that plant throughput can be achieved. (author). 1 fig

  5. New Digital Metal-Oxide (MOx Sensor Platform

    Directory of Open Access Journals (Sweden)

    Daniel Rüffer

    2018-03-01

    Full Text Available The application of metal oxide gas sensors in Internet of Things (IoT devices and mobile platforms like wearables and mobile phones offers new opportunities for sensing applications. Metal-oxide (MOx sensors are promising candidates for such applications, thanks to the scientific progresses achieved in recent years. For the widespread application of MOx sensors, viable commercial offerings are required. In this publication, the authors show that with the new Sensirion Gas Platform (SGP a milestone in the commercial application of MOx technology has been reached. The architecture of the new platform and its performance in selected applications are presented.

  6. An evaluation of the deployment of AIROX-recycled fuel in pressurized water reactors

    International Nuclear Information System (INIS)

    Jahshan, S.N.; McGeehan, T.J.

    1994-01-01

    An analytical evaluation is made of the pressurized water reactor (PWR) in-core performance of recycled light water reactor fuel that has been Atomics International reduction oxidation (AIROX) reprocessed and reenriched with fissile materials. The neutronics performance is shown to lie within the neutronics performance of existing high-performance and high-burnup fuels. Three AIROX-recycled fuels are compared with a high-burnup virgin fuel and an equivalent mixed-oxide (MOX) fuel. The AIROX-recycled fuel neutronics performance lies consistently between the virgin and the MOX fuel for both the pin power peaking and the reactivity response characteristics in PWRs. Among the attractive features of AIROX-recycled fuel is that it can optimize fissile and fertile fuel use, minimize final fuel disposal impact on the environment, and provide energy in the process of denaturing weapons-grade fissile materials. The fuel material performance may be anticipated from high-burnup virgin fuel and from MOX fuel performance. Recommendations for lead rod testing and for optimization of the AIROX-processing and resintering techniques are made

  7. Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P.O. 1236909 Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

    1978-12-01

    The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel

  8. Effect of long-term storage of LWR spent fuel on Pu-thermal fuel cycle

    International Nuclear Information System (INIS)

    Kurosawa, Masayoshi; Naito, Yoshitaka; Suyama, Kenya; Itahara, Kuniyuki; Suzuki, Katsuo; Hamada, Koji

    1998-01-01

    According to the Long-term Program for Research, Development and Utilization of Nuclear Energy (June, 1994) in Japan, the Rokkasho Reprocessing Plant will be operated shortly after the year 2000, and the planning of the construction of the second commercial plant will be decided around 2010. Also, it is described that spent fuel storage has a positive meaning as an energy resource for the future utilization of Pu. Considering the balance between the increase of spent fuels and the domestic reprocessing capacity in Japan, it can be expected that the long-term storage of UO 2 spent fuels will be required. Then, we studied the effect of long-term storage of spent fuels on Pu-thermal fuel cycle. The burnup calculation were performed on the typical Japanese PWR fuel, and the burnup and criticality calculations were carried out on the Pu-thermal cores with MOX fuel. Based on the results, we evaluate the influence of extending the spent fuel storage term on the criticality safety, shielding design of the reprocessing plant and the core life time of the MOX core, etc. As the result of this work on long-term storage of LWR spent fuels, it becomes clear that there are few demerits regarding the lifetime of a MOX reactor core, and that there are many merits regarding the safety aspects of the fuel cycle facilities. Furthermore, long-term storage is meaningful as energy storage for effective utilization of Pu to be improved by technological innovation in future, and it will allow for sufficient time for the important policymaking of nuclear fuel cycle establishment in Japan. (author)

  9. Characterization and heading of irradiated fuels and their chemical analogs

    International Nuclear Information System (INIS)

    Serrano, J. A.

    2000-01-01

    This work presents results of leaching experiments under deionized water and under synthetic granite at room temperature in air using spent fuel (UO 2 and MOX LWR fuels) and the chemical analogues, natural UO 2 and SIMFUEL. The experimental conditions and procedure for irradiated and non-irradiated materials were kept similar as much as possible. Also dissolution behaviour studies of preoxidised LWR UO 2 and MOX spent fuel up to different on the oxidation degree. For both fuel types, UO 2 and MOX, the fission products considered showed a fractional release normalised to uranium higher than 1, due to either the larger inventory at preferential leaching zones, such as, grain boundaries or to the inherent higher solubility of some of these elements. In contrast to fission products, the fractional release of PU from the UO 2 fuel was not affected by the oxidation level. Finally a thermodynamic study of the experimental leaching results obtained in this work was performed. (Author)

  10. Nuclear fuel behaviour modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2001-07-01

    The Technical Committee Meeting (TCM) included separate sessions on the specific topics of fuel thermal performance and fission product retention. On thermal performance, it is apparent that the capability exists to measure conductivity in high burnup fuel either by out-of-pile measurement or by instrumentation of test reactor rods. State-of-the-art modelling codes contain models for the conductivity degradation process, and hence adequate predictions of fuel temperature are achievable. Concerning fission product release, it is clear that many groups around the world are actively investigating the subject, with experimental and modelling programmes being pursued. However, a general consensus on the exact mechanisms of gas release and related gas bubble swelling has yet to emerge, even at medium burnup levels. Fission gas phenomena, not only the release to open volumes, but the whole sequence of processes taking place prior to this, need to be modelled in any modern fuel performance code. The presence of gaseous fission products may generate rapid fuel swelling during power transients, and this can cause PCI and rod failure. At high burnups, the quantity of released gases could give rise to pressures exceeding the safe limits. Modelling of pellet-cladding interaction (PCI) effects during transient operation is also an active area of study for many groups. In some situations a purely empirical approach to failure modelling can be justified, while for other applications a more detailed mechanistic approach is required. Another aspect of cladding modelling which was featured at the TCM concerned corrosion and hydriding. Although this issue can be the main life-limiting factor on fuel duty, it is apparent that modelling methods, and the experimental measurement techniques that underpin them, are adequate. A session was included on MOX fuel modelling. Substantial programmes of work, especially by the MOX vendors, appear to be underway to bring the level of understanding

  11. 3-D extension C5G7 MOX benchmark calculation using threedant code

    International Nuclear Information System (INIS)

    Kim, H.Ch.; Han, Ch.Y.; Kim, J.K.; Na, B.Ch.

    2005-01-01

    It pursued the benchmark on deterministic 3-D MOX fuel assembly transport calculations without spatial homogenization (C5G7 MOX Benchmark Extension). The goal of this benchmark is to provide a more through test results for the abilities of current available 3-D methods to handle the spatial heterogeneities of reactor core. The benchmark requires solutions in the form of normalized pin powers as well as the eigenvalue for each of the control rod configurations; without rod, with A rods, and with B rods. In this work, the DANTSYS code package was applied to analyze the 3-D Extension C5G7 MOX Benchmark problems. The THREEDANT code within the DANTSYS code package, which solves the 3-D transport equation in x-y-z, and r-z-theta geometries, was employed to perform the benchmark calculations. To analyze the benchmark with the THREEDANT code, proper spatial and angular approximations were made. Several calculations were performed to investigate the effects of the different spatial approximations on the accuracy. The results from these sensitivity studies were analyzed and discussed. From the results, it is found that the 4*4 grid per pin cell is sufficiently refined so that very little benefit is obtained by increasing the mesh size. (authors)

  12. Quantitative dissolution of (U, Pu)O2 MOX (0.4% to 44% PuO2) using microwave heating technique

    International Nuclear Information System (INIS)

    Malav, R.K.; Fulzele, Ajit K.; Prakash, Amrit; Afzal, Md.; Panakkal, J.P.

    2011-01-01

    AFFF has fabricated the (U, Pu)O 2 mixed oxide fuels for PHWRs, BWRs, PFBRs and FBTRs. The quantitative dissolution of the fuel samples are required within time for accurate determination of uranium-plutonium in chemical quality control laboratory. This paper describes the use of microwave heating technique in quantitative dissolution of (U, Pu)O 2 MOX (from 0.4% to 44% PuO 2 ). (author)

  13. Development of advanced mixed oxide fuels for plutonium management

    International Nuclear Information System (INIS)

    Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

    1997-01-01

    A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium. (author)

  14. Development of advanced mixed oxide fuels for plutonium management

    International Nuclear Information System (INIS)

    Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

    1997-06-01

    A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium

  15. Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

    International Nuclear Information System (INIS)

    Pavlovitchev, A.M.

    2000-01-01

    The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes

  16. Fuel assembly

    International Nuclear Information System (INIS)

    Fushimi, Atsushi; Shimada, Hidemitsu; Aoyama, Motoo; Nakajima, Junjiro

    1998-01-01

    In a fuel assembly for an n x n lattice-like BWR type reactor, n is determined to 9 or greater, and the enrichment degree of plutonium is determined to 4.4% by weight or less. Alternatively, n is determined to 10 or greater, and the enrichment degree of plutonium is determined to 5.2% by weight or less. An average take-out burnup degree is determined to 39GWd/t or less, and the matrix is determined to 9 x 9 or more, or the average take-out burnup degree is determined to 51GWd/t, and the matrix is determined to 10 x 10 or more and the increase of the margin of the maximum power density obtained thereby is utilized for the compensation of the increase of distortion of power distribution due to decrease of the kinds of plutonium enrichment degree, thereby enabling to reduce the kind of the enrichment degree of MOX fuel rods to one. As a result, the manufacturing step for fuel pellets can be simplified to reduce the manufacturing cost for MOX fuel assemblies. (N.H.)

  17. Nuclear spent fuel management scenarios. Status and assessment report

    International Nuclear Information System (INIS)

    Dufek, J.; Arzhanov, V.; Gudowski, W.

    2006-06-01

    The strategy for management of spent nuclear fuel from the Swedish nuclear power programme is interim storage for cooling and decay for about 30 years followed by direct disposal of the fuel in a geologic repository. In various contexts it is of interest to compare this strategy with other strategies that might be available in the future as a result of ongoing research and development. In particular partitioning and transmutation is one such strategy that is subject to considerable R and D-efforts within the European Union and in other countries with large nuclear programmes. To facilitate such comparisons for the Swedish situation, with a planned phase out of the nuclear power programme, SKB has asked the team at Royal Inst. of Technology to describe and explore some scenarios that might be applied to the Swedish programme. The results of this study are presented in this report. The following scenarios were studied by the help of a specially developed computer programme: Phase out by 2025 with direct disposal. Burning plutonium and minor actinides as MOX in BWR. Burning plutonium and minor actinides as MOX in PWR. Burning plutonium and minor actinides in ADS. Combined LWR-MOX plus ADS. For the different scenarios nuclide inventories, waste amounts, costs, additional electricity production etc have been assessed. As a general conclusion it was found that BWR is more efficient for burning plutonium in MOX fuel than PWR. The difference is approximately 10%. Furthermore the BWR produces about 10% less americium inventory. An ADS reactor park can theoretically in an ideal case burn (transmute) 99% of the transuranium isotopes. The duration of such a scenario heavily depends on the interim time needed for cooling the spent fuel before reprocessing. Assuming 10 years for cooling of nuclear fuel from ADS, the duration will be at least 200 years under optimistic technical assumptions. The development and use of advanced pyro-processing with an interim cooling time of only

  18. Nuclear spent fuel management scenarios. Status and assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Dufek, J.; Arzhanov, V.; Gudowski, W. [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Nuclear and Reactor Physics

    2006-06-15

    The strategy for management of spent nuclear fuel from the Swedish nuclear power programme is interim storage for cooling and decay for about 30 years followed by direct disposal of the fuel in a geologic repository. In various contexts it is of interest to compare this strategy with other strategies that might be available in the future as a result of ongoing research and development. In particular partitioning and transmutation is one such strategy that is subject to considerable R and D-efforts within the European Union and in other countries with large nuclear programmes. To facilitate such comparisons for the Swedish situation, with a planned phase out of the nuclear power programme, SKB has asked the team at Royal Inst. of Technology to describe and explore some scenarios that might be applied to the Swedish programme. The results of this study are presented in this report. The following scenarios were studied by the help of a specially developed computer programme: Phase out by 2025 with direct disposal. Burning plutonium and minor actinides as MOX in BWR. Burning plutonium and minor actinides as MOX in PWR. Burning plutonium and minor actinides in ADS. Combined LWR-MOX plus ADS. For the different scenarios nuclide inventories, waste amounts, costs, additional electricity production etc have been assessed. As a general conclusion it was found that BWR is more efficient for burning plutonium in MOX fuel than PWR. The difference is approximately 10%. Furthermore the BWR produces about 10% less americium inventory. An ADS reactor park can theoretically in an ideal case burn (transmute) 99% of the transuranium isotopes. The duration of such a scenario heavily depends on the interim time needed for cooling the spent fuel before reprocessing. Assuming 10 years for cooling of nuclear fuel from ADS, the duration will be at least 200 years under optimistic technical assumptions. The development and use of advanced pyro-processing with an interim cooling time of only

  19. Fuel manufacturing and utilization

    International Nuclear Information System (INIS)

    2005-01-01

    The efficient utilisation of nuclear fuel requires manufacturing facilities capable of making advanced fuel types, with appropriate quality control. Once made, the use of such fuels requires a proper understanding of their behaviour in the reactor environment, so that safe operation for the design life can be achieved. The International Atomic Energy Agency supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle. It provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection. The IAEA supports the development fuel modelling expertise in Member States, covering both normal operation and postulated and severe accident conditions. It provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation. The IAEA supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, it provides information and support research into the basic properties of fuel materials, including UO 2 , MOX and zirconium alloys. It further offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology

  20. Key points for the design of Mox facilities

    International Nuclear Information System (INIS)

    Ducroux, R.; Gaiffe, L.; Dumond, S.; Cret, L.

    1998-01-01

    The design of a MOX fuel fabrication facility involves specific technical difficulties: - Process aspects: for example, its is necessary to meet the stringent requirements on the end products, while handling large quantities of powders and pellets; - Safety aspects: for example, containment of radioactive materials requires to use gloveboxes, to design process equipment so as to limit dispersion to the gloveboxes and to use systems for dust collection. - Technological aspects: for example, it is necessary to take into account maintenance early in the design, in order to lower the operation costs and lower the dose to the personnel. - Quality control and information systems: for example, it is necessary to be able to trace all the different products (powder lots, pellets, rods, assemblies). The design methods and organization set-up by COGEMA enables to master these technical difficulties during the different design steps and to obtain a MOX fabrication facility at the best performance versus cost compromise. These design methods rely mainly on: - taking into account all the different above mentioned constraints from the very beginning of the design process (by using the know-how resulting from experience feed-back, and also specific design tools developed by COGEMA and SGN); - launching a technical development and testing program at the beginning of the project and incorporating its results in the course of the design. (author)

  1. Assessment of Startup Fuel Options for a Test or Demonstration Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, Jon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Walters, L. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO2 and UO2-PuO2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availability are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.

  2. Fuel reprocessing/fabrication interface

    International Nuclear Information System (INIS)

    Benistan, G.; Blanchon, T.; Galimberti, M.; Mignot, E.

    1987-01-01

    EDF has conducted a major research, development and experimental programme concerning the recycling of plutonium and reprocessed uranium in pressurized water reactors, in collaboration with its major partners in the nuclear fuel cycle industry. Studies already conducted have demonstrated the technical and economic advantages of this recycling, as also its feasibility with due observance of the safety and reliability criteria constantly applied throughout the industrial development of the nuclear power sector in France. Data feedback from actual experience will make it possible to control the specific technical characteristics of MOX and reprocessed uranium fuels to a higher degree, as also management, viewed from the economic standpoint, of irradiated fuels and materials recovered from reprocessing. The next step will be to examine the reprocessing of MOX for reprocessed uranium fuels, either for secondary recycling in the PWR units, or, looking further ahead, in the fast breeders or later generation PWR units, after a storage period of a few years

  3. 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)

  4. Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

    Directory of Open Access Journals (Sweden)

    Asah-Opoku Fiifi

    2014-01-01

    Full Text Available Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX, uranium oxide fuel (UOX, and commercially enriched uranium or uranium metal (CEU - are used in this simulation and their impact on the effective multiplication factor (Keff and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

  5. 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)

  6. Analysis of transition to fuel cycle system with continuous recycling in fast and thermal reactors - 5060

    International Nuclear Information System (INIS)

    Passereini, S.; Feng, B.; Fei, T.; Kim, T.K.; Taiwo, T.A.; Brown, N.R.; Cuadra, A.

    2015-01-01

    A recent Evaluation and Screening study of nuclear fuel cycle options identified a few groups of options as most promising. One of these most promising Evaluation Groups (EGs) is characterized by the continuous recycling of uranium (U) and transuranics (TRU) with natural uranium feed in both fast and thermal critical reactors. This evaluation group, designated as EG30, is represented by an example fuel cycle option that employs a two-technology, two-stage fuel cycle system. The first stage involves the continuous recycling of co-extracted U/TRU in Sodium-cooled Fast Reactors (SFRs) with metallic fuel and breeding ratio greater than 1. The second stage involves the use of the surplus TRU in Mixed Oxide (MOX) fuel in Pressurized Water Reactors that are MOX-capable (MOX-PWRs). This paper presents and discusses preliminary fuel cycle analysis results from the fuel cycle codes VISION and DYMOND for the transition to this fuel cycle option from the current once-through cycle in the United States (U.S.) that consists of Light Water Reactors (LWRs) that only use conventional UO 2 fuel. The analyses in this paper are applicable for a constant 100 GWe capacity, roughly the size of the U.S. nuclear fleet. Two main strategies for the transition to EG30 were analyzed: 1) deploying both SFRs and MOX-PWRs in parallel or 2) deploying them in series with the SFR fleet first. With an estimated retirement schedule for the existing LWRs, an assumed reactor lifetime of 60 years, and no growth, the nuclear system fully transitions to the new fuel cycle within 100 years for both strategies without SFR fuel shortages. Compared to the once-through cycle, transition to the SFR/MOX-PWR fleet with continuous recycle was shown to offer significant reductions in uranium consumption and waste disposal requirements. In addition, these initial calculations revealed a few notable modeling and strategy questions regarding how recycled resources are allocated, reactors that can switch between

  7. Chemical characterisation of MOX grinder sludge and process evaluation for its dry recycling

    Energy Technology Data Exchange (ETDEWEB)

    Mallik, G K; Fulzele, A K; Kothari, M; Bhargava, V K; Kamath, H S [Bhabha Atomic Research Centre, Tarapur (India). Advanced Fuel Fabrication Facility

    1997-09-01

    A large quantity of sludge (approximately 5%) is generated as a result of centreless grinding of MOX pellets. Plutonium and uranium are recovered from such sludge, consisting of coolant, resin and some metallic impurities, by a wet chemical route. A case has been made for the recycling of the sludge by an optimum dry route on the basis of chemical characterisation of sludge generated at Advanced Fuel Fabrication Facility using diamond grinding wheel. (author). 2 tabs.

  8. Chemical characterisation of MOX grinder sludge and process evaluation for its dry recycling

    International Nuclear Information System (INIS)

    Mallik, G.K.; Fulzele, A.K.; Kothari, M.; Bhargava, V.K.; Kamath, H.S.

    1997-01-01

    A large quantity of sludge (approximately 5%) is generated as a result of centreless grinding of MOX pellets. Plutonium and uranium are recovered from such sludge, consisting of coolant, resin and some metallic impurities, by a wet chemical route. A case has been made for the recycling of the sludge by an optimum dry route on the basis of chemical characterisation of sludge generated at Advanced Fuel Fabrication Facility using diamond grinding wheel. (author). 2 tabs

  9. Physics characteristics of a CANDU-600 with repositional adjuster rods fuelled with MOX or natural uranium

    International Nuclear Information System (INIS)

    Boczar, P.G.

    1985-06-01

    Repositioning the adjuster rods in 4 axial banks in future CANDU-600 reactors would permit the flexibility of grading the inner and outer banks to achieve optimal flattening of the axial power distribution for any particular fuel. With the 4 banks identical, acceptable axial power profiles can be achieved for both MOX and natural uranium fuels. Future work is to be directed at assessing the impact of lower zone controller and shutoff rod worth in the configuration of reactivity devices considered here, and if necessary, in identifying means of increasing their worth

  10. New Concept of Designing Composite Fuel for Fast Reactors with Closing Fuel Cycle

    International Nuclear Information System (INIS)

    Savchenko, A.; Vatulin, A.; Uferov, O.; Kulakov, G.; Sorokin, V.

    2013-01-01

    For fast reactors a novel type of promising composite U-PuO2 fuel is proposed which is based on dispersion fuel elements. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. Novel fuel features higher characteristics in comparison to metallic or MOX fuel its fabrication technology is readily accomplished and is environmentally clean. A possibility is demonstrated of fabricating coated steel claddings to protect from interaction with fuel and fission products when use standard rod type MOX or metallic U-Pu-Zr fuel. Novel approach to reprocessing of composite fuel is demonstrated, which allows to separate uranium from burnt plutonium as well as the newly generated fissile plutonium from burnt one without chemical processes, which simplifies the closing of the nuclear fuel cycle. Novel composite fuel combines the advantages of metallic and ceramic types of fuel and has high uranium density that allows also to implicate it in BREST types reactor with conversion ratio more than 1. Peculiarities of closing nuclear cycle with composite fuel are demonstrated that allows more effective re-usage of generated Pu as well as, minimizing r/a wastes by incineration of MA in specially developed IMF design

  11. Spent fuel management in France: Reprocessing, conditioning, recycling

    International Nuclear Information System (INIS)

    Giraud, J.P.; Montalembert, J.A. de

    1994-01-01

    The French energy policy has been based for 20 years on the development of nuclear power. The some 75% share of nuclear in the total electricity generation, representing an annual production of 317 TWh requires full fuel cycle control from the head-end to the waste management. This paper presents the RCR concept (Reprocessing, Conditioning, Recycling) with its industrial implementation. The long lasting experience acquired in reprocessing and MOX fuel fabrication leads to a comprehensive industrial organization with minimized impact on the environment and waste generation. Each 900 MWe PWR loaded with MOX fuel avoids piling up 2,500 m 3 per year of mine tailings. By the year 2000, less than 500 m 3 of high-level and long-lived waste will be annually produced at La Hague for the French program. The fuel cycle facilities and the associated MOX loading programs are ramping-up according to schedule. Thus, the RCR concept is a reality as well as a policy adopted in several countries. Last but not least, RCR represents a strong commitment to non-proliferation as it is the way to fully control and master the plutonium inventory

  12. Analysis of a Partial MOX Core Design with Tritium Targets for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anistratov, Dmitriy Y. [Texas A & M Univ., College Station, TX (United States); Adams, Marvin L. [Texas A & M Univ., College Station, TX (United States)

    1998-04-19

    This report constitutes tangible and verifiable deliverable associated with the task To study the effects of using WG MOX fuel in tritium-producing LWR” of the subproject Water Reactor Options for Disposition of Plutonium. The principal investigators of this subproject are Naeem M. Abdurrahman of the University of Texas at Austin and Marvin L. Adams of Texas A&M University. This work was sponsored by the Amarillo National Resource Center for Plutonium.

  13. Ultrasmall PdmMn1-mOx binary alloyed nanoparticles on graphene catalysts for ethanol oxidation in alkaline media

    Science.gov (United States)

    Ahmed, Mohammad Shamsuddin; Park, Dongchul; Jeon, Seungwon

    2016-03-01

    A rare combination of graphene (G)-supported pallad