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.)

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

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.

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

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

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

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

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.

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.)

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 ...

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)

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)

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)

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.

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.)

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

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

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.)

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

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.)

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)

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)

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

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)

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)

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)

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)

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.

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

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

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

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)

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)

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

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)

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)

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)

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.

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)

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)

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)

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

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.

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.

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

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

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)

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)

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

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

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

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

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)

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

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)

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.

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)

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)

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.)

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)

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)

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)

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)

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

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)

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)

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

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

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.)

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)

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)

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

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.

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

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)

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

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

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

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

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

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

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

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

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)

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)

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.

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

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)

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

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

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)

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

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

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.

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)

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.

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)

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)

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)

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

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)

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.

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

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).

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

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)

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)

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)

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)

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

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)

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)

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.)

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

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

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)

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

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.

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

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)

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)

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)

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.)

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

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)

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)

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.

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.

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.

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.)

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

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)

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

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

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)

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

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)

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)

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)

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

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.

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)

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

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)

BR2 Reactor: Irradiation of fuels

International Nuclear Information System (INIS)

Verwimp, A.

2005-01-01

Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

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)

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)

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)

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

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.

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

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)

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)

The fuel of nuclear reactors

International Nuclear Information System (INIS)

1995-03-01

This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

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

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.

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)

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.)

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

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.

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)

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)

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.

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

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)

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

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)

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'.

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)

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

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.

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)

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

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)

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.

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

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

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.)

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

Framatome steady-state fission gas release database includes more than 290 fuel rods irradiated in commercial and experimental reactors with rod average burnups up to 67 GWd/tM. The transient database includes close to 60 fuel rods with burnups up to 62 GWd//tM. The hold time for these rods ranged from several minutes to many hours and the linear heat generation rates ranged from 30 kW/m to 50 kW/m. The quality of the fission gas release model is state-of-the-art as the uncertainty of the model is comparable to other code models. Framatome is also greatly concerned with the MOX fuel performance and modeling given that, since 1997, more than 1500 MOX fuel assemblies have been delivered to French and foreign PWRs. The paper focuses on the significant data acquired through surveillance and analytical programs used for the validation and the improvement of the MOX fuel modeling. (author)

Fuel cycle options for light water reactors in Germany

International Nuclear Information System (INIS)

Broecking, D.; Mester, W.

1999-01-01

In Germany 19 nuclear power plants with an electrical output of 22 GWe are in operation. Annually about 450 t of spent fuel are unloaded from the reactors. Currently most of the spent fuel elements are shipped to France and the United Kingdom for reprocessing according to contracts which have been signed since the late 70es. By the amendment of the Atomic Energy Act in 1994 the previous priority for reprocessing of spent nuclear fuel was substituted by a legal equivalency of the reprocessing and direct disposal option. As a consequence some utilities take into consideration the direct disposal of their spent fuel for economical reasons. The separated plutonium will be recycled as MOX fuel in light water reactors. About 30 tons of fissile plutonium will be available to German utilities for recycling by the year 2000. Twelve German reactors are already licensed for the use of MOX fuel, five others have applied for MOX use. Eight reactors are currently using MOX fuel or used it in the past. The spent fuel elements which shall be disposed of without reprocessing will be stored in two interim dry storage facilities at Gorleben and Ahaus. The storage capacities are 3800 and 4200 tHM, respectively. The Gorleben salt dome is currently investigated to prove its suitability as a repository for high level radioactive waste, either in a vitrified form or as conditioned spent fuel. The future development of the nuclear fuel cycle and radioactive waste management depends on the future role of nuclear energy in Germany. According to estimations of the German utilities no additional nuclear power plants are needed in the near future. Around the middle of the next decade it will have to be decided whether existing plants should be substituted by new ones. For the foreseeable time German utilities are interested in a highly flexible approach to the nuclear fuel cycle and waste management keeping open both spent fuel management options: the closed fuel cycle and direct disposal of

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)

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

Characterization of aerosols from industrial fabrication of mixed-oxide nuclear reactor fuels

International Nuclear Information System (INIS)

Hoover, M.D.; Newton, G.J.

1997-01-01

Recycling plutonium into mixed-oxide (MOX) fuel for nuclear reactors is being given serious consideration as a safe and environmentally sound method of managing plutonium from weapons programs. Planning for the proper design and safe operation of the MOX fuel fabrication facilities can take advantage of studies done in the 1970s, when recycling of plutonium from nuclear fuel was under serious consideration. At that time, it was recognized that the recycle of plutonium and uranium in irradiated fuel could provide a significant energy source and that the use of 239 Pu in light water reactor fuel would reduce the requirements for enriched 235 U as a reactor fuel. It was also recognized that the fabrication of uranium and plutonium reactor fuels would not be risk-free. Despite engineered safety precautions such as the handling of uranium and plutonium in glove-box enclosures, accidental releases of radioactive aerosols from normal containment might occur. Workers might then be exposed to the released materials by inhalation

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.

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

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

Development, Fabrication and Characterization of Fuels for Indian Fast Reactor Programme

International Nuclear Information System (INIS)

Kumar, Arun

2013-01-01

Development of Fast Reactor fuels in India started in early Seventies. The successful development of Mixed Carbide fuels for FBTR and MOX fuel for PFBR have given confidence in manufacture of fuels for Fast Reactors. Effort is being put to develop high Breeding Ratio Metallic fuel (binary/ternary). Few fuel pins have been fabricated and is under test irradiation. However, this is only a beginning and complete fuel cycle activities are under development. Metal fuelled Fast Reactors will provide high growth rate in Indian Fast Reactor programme

A Simplified Supercritical Fast Reactor with Thorium Fuel

Directory of Open Access Journals (Sweden)

Peng Zhang

2014-01-01

Full Text Available Super-Critical water-cooled Fast Reactor (SCFR is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure and keeping negative coolant void reactivity during the whole core life. A core burnup simulation scheme based on Monte Carlo lattice homogenization is adopted in this study, and the reactor physics analysis has been performed with DU-MOX and Th-MOX fuel. The main issues discussed include the fuel conversion ratio and the coolant void reactivity. The analysis shows that thorium-based fuel can provide inherent safety for SCFR without use of blanket, which is favorable for the mechanical design of SCFR.

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

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)

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)

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)

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

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)

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)

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)

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

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)

Proliferation resistance of small modular reactors fuels

Energy Technology Data Exchange (ETDEWEB)

Polidoro, F.; Parozzi, F. [RSE - Ricerca sul Sistema Energetico,Via Rubattino 54, 20134, Milano (Italy); Fassnacht, F.; Kuett, M.; Englert, M. [IANUS, Darmstadt University of Technology, Alexanderstr. 35, D-64283 Darmstadt (Germany)

2013-07-01

In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

Characteristics of fast reactor core designs and closed fuel cycle

International Nuclear Information System (INIS)

Poplavsky, V.M.; Eliseev, V.A.; Matveev, V.I.; Khomyakov, Y.S.; Tsyboulya, A.M.; Tsykunov, A.G.; Chebeskov, A.N.

2007-01-01

On the basis of the results of recent studies, preliminary basic requirements related to characteristics of fast reactor core and nuclear fuel cycle were elaborated. Decreasing reactivity margin due to approaching breeding ratio to 1, requirements to support non-proliferation of nuclear weapons, and requirements to decrease amount of radioactive waste are under consideration. Several designs of the BN-800 reactor core have been studied. In the case of MOX fuel it is possible to reach a breeding ratio about 1 due to the use of larger size of fuel elements with higher fuel density. Keeping low axial fertile blanket that would be reprocessed altogether with the core, it is possible to set up closed fuel cycle with the use of own produced plutonium only. Conceptual core designs of advanced commercial reactor BN-1800 with MOX and nitride fuel are also under consideration. It has been shown that it is expedient to use single enrichment fuel core design in this reactor in order to reach sufficient flattening and stability of power rating in the core. The main feature of fast reactor fuel cycle is a possibility to utilize plutonium and minor actinides which are the main contributors to the long-living radiotoxicity in irradiated nuclear fuel. The results of comparative analytical studies on the risk of plutonium proliferation in case of open and closed fuel cycle of nuclear power are also presented in the paper. (authors)

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)

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

Application of powder metallurgy in production of nuclear fuels for research and power reactors

International Nuclear Information System (INIS)

Fukuda, Kosaku

2000-01-01

Powder metallurgy has been applied in many of the processes of nuclear fuel fabrication, which has contributed, to a great progress of the nuclear technology to date. Evolution of nuclear fuels still continues to meet various emerging demands in terms of enhanced safety, economical effectiveness, non-proliferation and environmental mitigation. This paper reviews recent progress of nuclear fuels of research and power reactors, in particular, focusing on the powder metallurgy application. First, the review is made on plate type fuels for research reactors, inter alia, silicide fuel which is prevailing worldwide from the viewpoint of non-proliferation. The relation between fabrication and irradiation behavior is also discussed. Next, oxide fuels including MOX are reviewed. Recent interests of UO 2 are directed toward large grain pellets and burnable absorber pellets, both of which arise from requirement of extended burnup. Finally, the MOX fuel for thermal reactors is reviewed. (author)

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)

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)

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.)

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)

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.

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

Advanced fuels for plutonium management in pressurized water reactors

International Nuclear Information System (INIS)

Vasile, A.; Dufour, Ph.; Golfier, H.; Grouiller, J.P.; Guillet, J.L.; Poinot, Ch.; Youinou, G.; Zaetta, A.

2003-01-01

Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1 . More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate

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)

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.

Economic Analysis of Symbiotic Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

International Nuclear Information System (INIS)

Williams, Kent Alan; Shropshire, David E.

2009-01-01

A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle

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)

Fabrication of cermet fuel for fast reactor

International Nuclear Information System (INIS)

Mishra, Sudhir; Kumar, Arun; Kutty, T.R.G.; Kamath, H.S.

2011-01-01

Mixed oxide (MOX) (U,Pu)O 2 , and metallic (U,Pu ,Zr) fuels are considered promising fuels for the fast reactor. The fuel cycle of MOX is well established. The advantages of the oxide fuel are its easy fabricability, good performance in the reactor and a well established reprocessing technology. However the problems lie in low thermal conductivity , low density of the fuel leading to low breeding ratio and consequently longer doubling time. The metallic fuel has the advantages of high thermal conductivity, higher metal density and higher coefficient of linear expansion. The higher coefficient of linear expansion is good from the safety consideration (negative reactivity factor). Because of higher metal density it offers highest breeding ratio and shortest doubling time. Metallic fuel disadvantages comprise large swelling at high burnup, fuel cladding interaction and lower margin between operating and melting temperature. The optimal solution may lie in cermet fuel (U, PuO 2 ), where PuO 2 is dispersed in U metal matrix and combines the favorable features of both the fuel types. The advantages of this fuel include high thermal conductivity, larger margin between melting and operating temperature, ability to retain fission product etc. The matrix being of high density metal the advantage of high breeding ratio is also maintained. In this report some results of fabrication of cermet pellet comprising of UO 2 /PuO 2 dispersed in U metal powder through classical powder metallurgy route and characterization are presented. (author)

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)

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

Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

International Nuclear Information System (INIS)

Fukaya, Y.; Okubo, T.; Uchikawa, S.

2008-01-01

The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the 241 Pu content in the initial fuel, and the decay heat mainly depends on 238 Pu and 244 Cm. The contribution of 244 Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from

Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

Energy Technology Data Exchange (ETDEWEB)

Fukaya, Y. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)], E-mail: fukaya.yuji@jaea.go.jp; Okubo, T.; Uchikawa, S. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)

2008-07-15

The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the {sup 241}Pu content in the initial fuel, and the decay heat mainly depends on {sup 238}Pu and {sup 244}Cm. The contribution of {sup 244}Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum

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

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

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)

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

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)

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)

Investigation on innovative water reactor for flexible fuel cycle (FLWR). (1) Conceptual design

International Nuclear Information System (INIS)

Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiko; Ohnuki, Akira; Iwamura, Takamichi

2005-01-01

A concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI) in order to ensure sustainable energy supply in the future based on the well-experienced Light Water Reactor (LWR). The concept aims at effective and flexible utilization of uranium and plutonium resources through plutonium multiple recycling by two stages. In the first stage, the FLWR core 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 core in the second stage 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 core concepts in both stages utilize the compatible and the same size fuel assemblies, and hence during the reactor operation period, the former concept can proceed to the latter in the same reactor system, corresponding flexibly to the expected change in the future circumstances of natural uranium resource, or establishment of economical reprocessing technology of MOX spent fuel. The FLWR is essentially a BWR-type reactor, and its core design is characterized by use of hexagonal-shaped fuel assemblies with the triangular-lattice fuel rod configuration of highly enriched MOX fuel, control rods with Y-shaped blades, and a short and flat core design. Detailed investigations have been performed on the core design, in conjunction with the other related studies such as on thermal hydraulics in the tight lattice core including experimental activities, and the results obtained so far have shown the proposed concept is feasible and promising. (author)

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

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)

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

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)

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)

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)

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)

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)

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

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)

Reactor based plutonium disposition - physics and fuel behaviour benchmark studies of an OECD/NEA experts group

International Nuclear Information System (INIS)

D'Hondt, P.; Gehin, J.; Na, B.C.; Sartori, E.; Wiesenack, W.

2001-01-01

One of the options envisaged for disposing of weapons grade plutonium, declared surplus for national defence in the Russian Federation and Usa, is to burn it in nuclear power reactors. The scientific/technical know-how accumulated in the use of MOX as a fuel for electricity generation is of great relevance for the plutonium disposition programmes. An Expert Group of the OECD/Nea is carrying out a series of benchmarks with the aim of facilitating the use of this know-how for meeting this objective. This paper describes the background that led to establishing the Expert Group, and the present status of results from these benchmarks. The benchmark studies cover a theoretical reactor physics benchmark on a VVER-1000 core loaded with MOX, two experimental benchmarks on MOX lattices and a benchmark concerned with MOX fuel behaviour for both solid and hollow pellets. First conclusions are outlined as well as future work. (author)

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

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)

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

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.)

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

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

Advanced CANDU reactors fuel analysis through optimal fuel management at approach to refuelling equilibrium

International Nuclear Information System (INIS)

Tingle, C.P.; Bonin, H.W.

1999-01-01

The analysis of alternate CANDU fuels along with natural uranium-based fuel was carried out from the view point of optimal in-core fuel management at approach to refuelling equilibrium. The alternate fuels considered in the present work include thorium containing oxide mixtures (MOX), plutonium-based MOX, and Pressurised Water Reactor (PWR) spent fuel recycled in CANDU reactors (Direct Use of spent PWR fuel in CANDU (DUPIC)); these are compared with the usual natural UO 2 fuel. The focus of the study is on the 'Approach to Refuelling Equilibrium' period which immediately follows the initial commissioning of the reactor. The in-core fuel management problem for this period is treated as an optimization problem in which the objective function is the refuelling frequency to be minimized by adjusting the following decision variables: the channel to be refuelled next, the time of the refuelling and the number of fresh fuel bundles to be inserted in the channel. Several constraints are also included in the optimisation problem which is solved using Perturbation Theory. Both the present 37-rod CANDU fuel bundle and the proposed CANFLEX bundle designs are part of this study. The results include the time to reach refuelling equilibrium from initial start-up of the reactor, the average discharge burnup, the average refuelling frequency and the average channel and bundle powers relative to natural UO 2 . The model was initially tested and the average discharge burnup for natural UO 2 came within 2% of the industry accepted 199 MWh/kgHE. For this type of fuel, the optimization exercise predicted the savings of 43 bundles per full power year. In addition to producing average discharge burnups and other parameters for the advanced fuels investigated, the optimisation model also evidenced some problem areas like high power densities for fuels such as the DUPIC. Perturbation Theory has proven itself to be an accurate and valuable optimization tool in predicting the time between

Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

Energy Technology Data Exchange (ETDEWEB)

NONE

2009-06-15

SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

International Nuclear Information System (INIS)

2009-06-01

SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing facilities. - 3. Advances in Water

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)

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.

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

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)

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.

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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)

Studies on the safety and transmutation behaviour of innovative fuels for light water reactors

International Nuclear Information System (INIS)

Schitthelm, Oliver

2012-01-01

Nuclear power plants contribute a substantial part to the energy demand in industry. Today the most common fuel cycle uses enriched uranium which produces plutonium due to its 238 U content. With respect to the long-term waste disposal Plutonium is an issue due to its heat production and radiotoxicity. This thesis consists of three main parts. In the first part the development and validation of a new code package MCBURN for spatial high resolution burnup simulations is presented. In the second part several innovative uranium-free and plutonium-burning fuels are evaluated on assembly level. Candidates for these fuels are a thorium/plutonium fuel and an inert matrix fuel consisting of plutonium dispersed in an enriched molybdenum matrix. The performance of these fuels is evaluated against existing MOX and enriched uranium fuels considering the safety and transmutation behaviour. The evaluation contains the boron efficiency, the void coefficient, the doppler coefficient and the net balances of every radionuclide. In the third part these innovative fuels are introduced into a German KONVOI reactor core. Considering todays approved usage of MOX fuels a partial loading of one third of innovative fuels and two third of classical uranium fuels was analysed. The efficiency of the plutonium depletion is determined by the ratio of the production of higher isotopes compared to the plutonium depletion. Todays MOX-fuels transmutate about 25% to 30% into higher actinides as Americium or Curium. In uranium-free fuels this ratio is about 10% due to the lack of additional plutonium production. The analyses of the reactor core have shown that one third of MOX fuel is not capable of a net reduction of plutonium. On the other hand a partial loading with thorium/plutonium fuel incinerates about half the amount of plutonium produced by an uranium only core. If IMF is used the ratio increases to about 75%. Considering the safety behavior all fuels have shown comparable results.

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)

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

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)

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

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)

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

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)

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

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

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

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)

Analysis of Accident Scenarios for the Development of Probabilistic Safety Assessment Model for the Metallic Fuel Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Kim, Tae Woon; Park, S. Y.; Yang, J. E.; Kwon, Y. M.; Jeong, H. Y.; Suk, S. D.; Lee, Y. B.

2009-03-01

The safety analysis reports which were reported during the development of sodium cooled fast reactors in the foreign countries are reviewed for the establishment of Probabilistic Safety Analysis models for the domestic SFR which are under development. There are lots of differences in the safety characteristics between the mixed oxide (MOX) fuel SFR and metallic fuel SFR. Metallic fuel SFR is under development in Korea while MOX fuel SFR is under development in France, Japan, India and China. Therefore the status on the development of fast reactors in the foreign countries are reviewed at first and then the safety characteristics between the MOX fuel SFR and the metallic fuel SFR are reviewed. The core damage can be defined as coolant voiding, fuel melting, cladding damage. The melting points of metallic fuel and the MOX fuel is about 1000 .deg. C and 2300 .deg. C, respectively. The high energy stored in the MOX fuel have higher potential to voiding of coolant compared to the possibility in the metallic fuel. The metallic fuel has also inherent reactivity feedback characteristic that the metallic fuel SFR can be shutdown safely in the events of transient overpower, loss of flow, and loss of heat sink without scram. The metallic fuel has, however, lower melting point due to the eutectic formation between the uranium in metallic fuel and the ferrite in metallic cladding. It is needed to identify the core damage accident scenarios to develop Level-1 PSA model. SSC-K computer code is used to identify the conditions in which the core damage can occur in the KALIMER-600 SFR. The accident cases which are analyzed are the triple failure accidents such as unprotected transient over power events, loss of flow events, and loss of heat sink events with impaired safety systems or functions. Through the analysis of the triple failure accidents for the KALIMER-600 SFR, it is found that the PSA model developed for the PRISM reactor design can be applied to KALIMER-600. However

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)

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)

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)

Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

International Nuclear Information System (INIS)

Čerba, Štefan; Vrban, Branislav; Lüley, Jakub; Dařílek, Petr; Zajac, Radoslav; Nečas, Vladimír; Haščik, Ján

2014-01-01

Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice

Reactor physics experiment plan using TCA

Energy Technology Data Exchange (ETDEWEB)

Shimada, Shoichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-06-01

The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors, which aims 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. This report is to plan critical experiments using TCA in JAERI. Critical Experiments performed so far in Europe and Japan are 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 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 is obtained by calculations and modification of the equipment for the experiments are shown. New MOX fuel and UO{sub 2} fuel rods are necessary for the RMWR critical experiments. Number of MOX fuel rods is 1000 for Plutonium fissile enrichment of 5 wt%, 1000 for 10 wt%, 1500 for 15 wt% and 500 for 20 wt%, respectively. Depleted UO{sub 2} fuel rods for blanket/buffer region are 4000. Driver fuel rods of 4.9 wt% UO{sub 2} are 3000. Modification of TCA facility is requested to treat the large amount of MOX fuels from safety point of view. Additional shielding device at the top of the tank for loading the MOX fuels and additional safety plates to ensure safety are requested. The core is divided into two regions by inserting an inner tank to avoid criticality in MOX region only. The test region is composed by MOX fuel rods in the inner tank. Criticality is established by UO{sub 2} driver fuel rods outside of the inner tank. (Tsuchihashi, K.)

Using NJOY99 and MCNP4B2 to Estimate the Radiation Damage Displacements per Atom per Second in Steel Within the Boiling Water Reactor Core Shroud and Vessel Wall from Reactor-Grade Mixed-Oxide/Uranium Oxide Fuel for the Nuclear Power Plant at Laguna Verde, Veracruz, Mexico

International Nuclear Information System (INIS)

Vickers, Lisa

2003-01-01

The government of Mexico has expressed interest in utilizing the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18 to 30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons.There is concern that a core with a fraction of MOX fuel (i.e., increased 239 Pu wt%) would increase the radiation damage displacements per atom per second (dpa-s -1 ) in steel within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation damage within the core shroud and vessel wall is a concern because of the potentially adverse affect to personnel and public safety, environment, and operating life of the reactor.The primary uniqueness of this paper is the computation of radiation damage (dpa-s -1 ) using NJOY99-processed cross sections for steel within the core shroud and vessel wall. Specifically, the unique radiation damage results are several orders of magnitude greater than results of previous works. In addition, the conclusion of this paper was that the addition of the maximum fraction of one-third MOX fuel to the LV1 BWR core did significantly increase the radiation damage in steel within the core shroud and vessel wall such that without mitigation of radiation damage by periodic thermal annealing or reduction in operating parameters such as neutron fluence, core temperature, and pressure, it posed a potentially adverse affect to the personnel and public safety, environment, and operating life of the reactor

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

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)

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)

The integration of fast reactor to the fuel cycle in Slovakia

International Nuclear Information System (INIS)

Zajac, R.; Darilek, P.; Necas, V.

2009-01-01

A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of WWER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual WWER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the WWER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in WWER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from WWER-440 in the fission products. The next step is multi recycling of Pu in the fission products to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (Authors)

The integration of fast reactor to the fuel cycle in Slovakia

International Nuclear Information System (INIS)

Zajac, R.; Darilek, P.; Necas, V.

2009-01-01

A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of VVER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual VVER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the VVER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in VVER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from VVER-440 in the FR. The next step is multirecycling of Pu in the FR to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (authors)

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)

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

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)

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

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.

Fuel clad chemical interactions in fast reactor MOX fuels

Energy Technology Data Exchange (ETDEWEB)

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