Overview of the Vercors programme devoted to safety studies on irradiated PWR fuel

International Nuclear Information System (INIS)

Tourasse, M.; Andre, B.; Ducros, G.; Maro, D.

1996-01-01

The first objective of the Heva-Vercors Program is to improve the data base of fission product release and behaviour after an extensive fuel temperature increase and loss of integrity of the fuel elements that occur in case of severe PWR accident. The program is co-funded by the French Nuclear Protection and Safety Institute (IPSN) and Electricite de France (EdF). The experiments are conducted in a shielded cell of the French Grenoble Nuclear Centre. For these tests, industrial fuel from French PWR reactor plants is used. In order to rebuild the short lived fission product inventory, a reirradiation is performed in the experimental Siloe reactor, prior to the test. Eight tests have been conducted in the frame of the Heva Program up to 2370 K in the 1983-1988 period. The main outcomes of these studies were linked to the volatile fission product release. This program has been extended by the Vercors one with higher fuel temperature (2600 K) and improved instrumentation : gamma spectrometry, emission tomography, metallography, scanning electron microscopy, energy dispersive X-ray analysis, X-ray diffraction are some of the experimental techniques used for on-line and post-test characterization. The knowledge of the behaviour of low volatile fission product has been significantly improved with the six Vercors tests. The results of the Vercors 4 test (38 GWd/t(U), 2570 K, reducing atmosphere) are presented here as an example. The key parameters are exhibited. The next step of these studies will use the Vercors HT loop that is planned to be operational at the beginning of 1996 to reach fuel melting temperature. (author)

Gadolinia experience and design for PWR fuel cycles

International Nuclear Information System (INIS)

Stephenson, L. C.

2000-01-01

The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

EDF studies on PWR vessel internal loading

International Nuclear Information System (INIS)

Bellet, S.; Vallat, S.

1998-01-01

EDF has undertaken some mechanics and thermal-hydraulics studies with the objective of mastering plant phenomena today and in order to numerically predict the behaviour of vessel internals on units planned for the future. From some justifications already underway after in operation incidents (wear and drop time of RCCA rods, fuel deflection, adapter cracks, baffle bolt cracks) we intend to control reactor vessel flows and mechanical behaviour of internal structures. During normal operation, thermal-hydraulic is the main load of vessel internals. The current approach consists of acquiring the capacity to link different calculations, taking care that codes are qualified for physical phenomena and complex 3D geometries. For baffle assembly, a more simple model of this structure has been used to treat the physical phenomena linked to the LOCA transient. Results are encouraging mainly due to code capacity progression (resolution and models), which allows more and more complex physical phenomena to be treated, like turbulence flow and LOCA. (author)

Pushing back the boundaries of PWR fuel performance

International Nuclear Information System (INIS)

Sofer, G.A.; Skogen, F.B.; Brown, C.A.; Fresk, Y.U.

1985-01-01

In today's fiercely competitive PWR reload market utilities are benefiting from a variety of design innovations which are helping to cut fuel cycle costs and to improve fuel performance. An advanced PWR fuel design from Exxon, for example, currently under evaluation at the Ginna plant in the United States, offers higher burn-up and greater power cycling. (author)

An economic analysis code used for PWR fuel cycle

International Nuclear Information System (INIS)

Liu Dingqin

1989-01-01

An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

EDF energy generation UK transport of irradiated fuel

Energy Technology Data Exchange (ETDEWEB)

James, R. [EDF Energy, London, (United Kingdom)

2015-07-01

This paper give an overview of irradiated fuel transport in the UK. It describes the design of irradiated fuel flask used by EDF Energy; operational experience and good practices learnt from over 50 years of irradiated fuel transport. The AGRs can store approximately 9 months generation of spent fuel, hence the ability to transport irradiated fuel is vital. Movements are by road to the nearest railhead, typically less than 2 miles and then by rail to Sellafield, up to 400 miles, for reprocessing or long term storage. Road and rail vehicles are covered. To date in the UK: over 30,000 Magnox flask journeys and over 15,000 AGR A2 flask journeys have been carried out.

The safety approach in the operation of EDF power plants

International Nuclear Information System (INIS)

Bertron, L.; Mira, J.J.

1988-01-01

To get a view on what is involved in maintaining a high level of safety in the operation of EdF nuclear power plants, it may be recalled that in 1987, 76 % of the EdF production was nuclear. The nuclear plants include thirty-four standard PWR 900 plants, fourteen PWR 1300 plants, the 305 MW SENA PWR, the four 500 MW GCR: CHINON A3 plant, St-LAURENT A1 (390 MW), A2 (450 MW) and BUGEY 1 (540 MW), the 233 MW PHENIX fast breeder reactor and the CREYS-MALVILLE 1200 MW fast breeder reactor, now being prepared for a new startup after the 1987 incident. So the importance of a safe operation of this investment is considerable for EdF, which is the designer, owner, industrial architect and operator. According to the French regulations, EdF is responsible for the safe operation of its power plants. A considerable human component is also at stake, as the safe operation of plants implies all the personnel to varying degrees. There are 15,000 such employees, all of whom have to be trained, competent and motivated. The operation of this system for 340 reactor-years has to-date resulted in no incident of any significant impact on the environment. Right from the start, safety in operation has always been an essential and clearly stated priority. Among other lessons the Three-Mile Island and Chernobyl accidents have reinforced the conviction that the human factors, the man-machine interface, and the safety culture were determining elements. With forty-eigh PWR plants in service, the problem is to maintain safe operation of a system now running at cruising speed, but also including some units (particularly the GCRs) that must be prepared for decommissioning. In addition EDF has to demonstrate the safe operations of CREYS MALVILLE, fast breeder reactor

PWR fuel performance and future trend in Japan

International Nuclear Information System (INIS)

Kondo, Y.

1987-01-01

Since the first PWR power plant Mihama Unit 1 initiated its commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts to improve PWR technology. The results are already seen in significantly improved performance of 16 PWR plants now in operation. Mitsubishi Heavy Industries Ltd. (MHI) has been supplying them with nuclear fuel assemblies, which are over 5700. As the reliability of the current design fuel has been achieved, the direction of R and D on nuclear fuel has changed to make nuclear power more competitive to the other power generation methods. The most important R and D targets are the burnup extension, Gd contained fuel, Pu utilizatoin and the load follow capacility. (author)

Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

International Nuclear Information System (INIS)

O'Leary, P.M.; Pitts, M.L.

2000-01-01

Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers

Development and testing of the EDF-2 reactor fuel element

International Nuclear Information System (INIS)

Delpeyroux, P.

1964-01-01

This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [fr

Highlights of the French program on PWR fuel

Energy Technology Data Exchange (ETDEWEB)

Pages, J P [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

1997-12-01

The presentation reviews the French programme on PWR fuel including the overall results of the year 1996 for nuclear operation; fuel management and economy; French nuclear electricity generation sites; production of nuclear generated electricity; energy availability of the 900 and 1,300 Mw PWR units; average radioactive liquid releases excluding tritium per unit; plutonium recycling experience.

Power variation and frequency regulation. Adaptation of PWR plant possibilities to the network needs

International Nuclear Information System (INIS)

Baboulin, J.P.; Burger, M.

1980-01-01

When the PWR are an important part of the power installed on a network, and that will be the case of the EDF network in the coming years, the participation of those plants to the power regulating becomes a necessity for the operating staff. This load regulating includes: daily variations of high amplitude; a permanent frequency - power regulating. The first part of the communication shows the network exploitation principles, and the resulting power variations concerning the existing nuclear power plants. Such transients are leading to stresses on fuel. The second part of the communication reports about the test program engaged by EDF in collaboration with the CEA and FRAMATOME, in order to study the fuel behaviour in real power conditions and power cycles, and that, just to the operational burn up of this fuel. (author)

Evolutionary developments of advanced PWR nuclear fuels and cladding materials

International Nuclear Information System (INIS)

Kim, Kyu-Tae

2013-01-01

Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10 −6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

Overview of the Vercors Programme Devoted to Safety Studies on Irradiated PWR Fuel

International Nuclear Information System (INIS)

Tourasse, M.; Andre, B.; Ducros, G.; Maro, D.

1996-01-01

The first objective of the Heva-Vercors Program is to improve the data of fission product release and behaviour after an extensive fuel temperature increase and loss of integrity of the fuel elements that occur in case of severe PWR accident. The program is co-funded by the French Nuclear Protection and Safety Institute (IPSN) and Electricite de France (EDF). The experiments are conducted in a shielded cell of the French Grenoble Nuclear Centre. For these tests, industrial fuel from French PWR reactor plants is used. In order to rebuild the short lived fission product inventory, a reirradiation is performed in the experimental Siloe reactor, prior to the test. Eight tests have been conducted in the frame of the Heva Program up to 2370 K in the 1983-1988 period. The main outcomes of these studies were linked to the volatile fission product release. This program has been extended by the Vercors one with higher fuel temperature (2600 K) and improved instrumentation: gamma spectrometry, emission tomography, metallography, scanning electron microscopy, energy dispersive X-ray analysis, X-ray diffraction are some of the experimental techniques used for on line and post test characterization. The knowledge of the behavior of low volatile fission product has been significantly improved with the six Vercors tests. The results of the Vercors 4 test (38 GWd/t(U), 2570 K, reducing atmosphere) are presented here as an example. The key parameters are exhibited. The next step of these studies will use the Vercors HT loop that is planned to be operational at the beginning of 1996 to reach fuel melting temperature. The first aim of these future tests is to study the behaviour of non volatile and transuranic elements. An even more sophisticated instrumentation is implemented to reach the goal. The use of MOX fuel, the interaction between fission product aerosols and structural materials (Ag-In-Cd) and the fuel granulometry effect will be the next steps of the experimental program

Overview of the Vercors Programme Devoted to Safety Studies on Irradiated PWR Fuel

Energy Technology Data Exchange (ETDEWEB)

Tourasse, M.; Andre, B.; Ducros, G. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire

1996-12-31

The first objective of the Heva-Vercors Program is to improve the data of fission product release and behaviour after an extensive fuel temperature increase and loss of integrity of the fuel elements that occur in case of severe PWR accident. The program is co-funded by the French Nuclear Protection and Safety Institute (IPSN) and Electricite de France (EDF). The experiments are conducted in a shielded cell of the French Grenoble Nuclear Centre. For these tests, industrial fuel from French PWR reactor plants is used. In order to rebuild the short lived fission product inventory, a reirradiation is performed in the experimental Siloe reactor, prior to the test. Eight tests have been conducted in the frame of the Heva Program up to 2370 K in the 1983-1988 period. The main outcomes of these studies were linked to the volatile fission product release. This program has been extended by the Vercors one with higher fuel temperature (2600 K) and improved instrumentation: gamma spectrometry, emission tomography, metallography, scanning electron microscopy, energy dispersive X-ray analysis, X-ray diffraction are some of the experimental techniques used for on line and post test characterization. The knowledge of the behavior of low volatile fission product has been significantly improved with the six Vercors tests. The results of the Vercors 4 test (38 GWd/t(U), 2570 K, reducing atmosphere) are presented here as an example. The key parameters are exhibited. The next step of these studies will use the Vercors HT loop that is planned to be operational at the beginning of 1996 to reach fuel melting temperature. The first aim of these future tests is to study the behaviour of non volatile and transuranic elements. An even more sophisticated instrumentation is implemented to reach the goal. The use of MOX fuel, the interaction between fission product aerosols and structural materials (Ag-In-Cd) and the fuel granulometry effect will be the next steps of the experimental program

A scheme of better utilization of PWR spent fuels

International Nuclear Information System (INIS)

Chung, Bum Jin; Kang, Chang Soon

1991-01-01

The recycle of PWR spent fuels in a CANDU reactor, so called the tandem fuel cycle is investigated in this study. This scheme of utilizing PWR spent fuels will ease the shortage of spent fuel storage capacity as well as will improve the use of uranium resources. The minimum modification the design of present CANDU reactor is seeked in the recycle. Nine different fuel types are considered in this work and are classified into two categories: refabrication and reconfiguration. For refabrication, PWR spent fuels are processed and refabricated into the present 37 rod lattice structure of fuel bundle, and for reconfiguration, meanwhile, spent fuels are simply disassembled and rods are cut to fit into the present grid configuration of fuel bundle without refabrication. For each fuel option, the neutronics calculation of lattice was conducted to evaluate the allowable burn up and distribution. The fuel cycle cost of each option was also computed to assess the economic justification. The results show that most tandem fuel cycle option considered in this study are technically feasible as well as economically viable. (Author)

PWR fuel physico chemistry. Improvements of the Sage code to compute thermochemical balance in PWR fuel

International Nuclear Information System (INIS)

Garcia, P.; Baron, D.; Piron, J.P.

1993-02-01

A physicochemical survey of high burnup fuel has been undertaken in the context of a 3-party action (CEA Cadarache - EDF/DER - FRAMATOME). One of the tasks involved consists in adapting the SAGE code for assessment of the thermochemical equilibria of fission products in solution in the fuel matrix. This paper describes the first stage of this task. Even if other improvements are planned, the oxid oxygen potentials are yet properly reproduced for the simulated burnup. (authors). 63 figs., 4 tabs., 41 refs

The cycle of the nuclear fuel used in EDF power plants

International Nuclear Information System (INIS)

2011-11-01

This document briefly indicates the different stages of the nuclear fuel cycle, from the purchase of natural uranium to waste storage. It also indicates the main responsibilities of EDF regarding this fuel cycle (to secure supplies, to organise material transportation, to process and store used fuels and associated wastes). It presents the different associated processes: uranium extraction, purification and concentration, conversion or fluoridation, enrichment. It briefly describes the fuel assembly fabrication, and indicates the main uranium producers in the world. Other addressed steps are: the transportation of fuel assembly, fuel loading, and spent fuel management, the processing of spent fuel and radioactive wastes

On site PWR fuel inspection measurements for operational and design verification

International Nuclear Information System (INIS)

1996-01-01

The on-site inspection of irradiated Pressurized Water Reactor (PWR) fuel and Non-Fuel Bearing Components (NFBC) is typically limited to visual inspections during refuelings using underwater TV cameras and is intended primarily to confirm whether the components will continue in operation. These inspections do not normally provide data for design verification nor information to benefit future fuel designs. Japanese PWR utilities and Nuclear Fuel Industries Ltd. designed, built, and performed demonstration tests of on-site inspection equipment that confirms operational readiness of PWR fuel and NFBC and also gathers data for design verification of these components. 4 figs, 3 tabs

PWR Fuel licensing in France - from design to reprocessing: licensing of nuclear PWR fuel rod design to satisfy with criteria for normal and abnormal fuel operation in France

International Nuclear Information System (INIS)

Beraha, R.

1999-01-01

In this lecture are presented: French regulatory context; Current fuel management methods; Request from the french operator EdF; Most recent actions of the french Nuclear safety authority; Fuel assemblies deformations (impact of high burn-up; investigations during reactor's exploitation; control rods drop off times)

Status and future perspectives of PWR and comparing views on WWER fuel technology

International Nuclear Information System (INIS)

Weidinger, H.

2003-01-01

The main purpose of this paper is to give an overview on status and future perspectives of the Western PWR fuel technology. For easer understanding and correlating, some comparing views to the WWER fuel technology are provided. This overview of the PWR fuel technology of course can not go into the details of the today used designs of fuel, fuel rods and fuel assemblies. However, it tries to describe the today achieved capability of PWR fuel technology with regard to reliability, efficiency and safety

Study on advanced nuclear fuel cycle of PWR/CANDU synergism

International Nuclear Information System (INIS)

Xie Zhongsheng; Huo Xiaodong

2002-01-01

According to the concrete condition that China has both PWR and CANDU reactors, one of the advanced nuclear fuel cycle strategy of PWR/CANDU synergism ws proposed, i.e. the reprocessed uranium of spent PWR fuel was used in CANDU reactor, which will save the uranium resource, increase the energy output, decrease the quantity of spent fuels to be disposed and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, the transition from the natural uranium to the recycled uranium (RU) can be completed without any changes of the structure of reactor core and operation mode. Furthermore, because of the low radiation level of RU, which is acceptable for CANDU reactor fuel fabrication, the present product line of fuel elements of CANDU reactor only need to be shielded slightly, also the conditions of transportation, operation and fuel management need not to be changed. Thus this strategy has significant practical and economical benefit

Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

Energy Technology Data Exchange (ETDEWEB)

Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

2015-07-01

In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

Scope and procedures of fuel management for PWR nuclear power plant

International Nuclear Information System (INIS)

Yao Zenghua

1997-01-01

The fuel management scope of PWR nuclear power plant includes nuclear fuel purchase and spent fuel disposal, ex-core fuel management, in-core fuel management, core management and fuel assembly behavior follow up. A suit of complete and efficient fuel management procedures have to be created to ensure the quality and efficiency of fuel management work. The hierarchy of fuel management procedure is divided into four levels: main procedure, administration procedure, implement procedure and technic procedure. A brief introduction to the fuel management scope and procedures of PWR nuclear power plant are given

Dissolution process for advanced-PWR-type fuels

International Nuclear Information System (INIS)

Black, D.E.; Decker, L.A.; Pearson, L.G.

1979-01-01

The new Fluorinel Dissolution Process and Fuel Storage (FAST) Facility at ICPP will provide underwater storage of spent PWR fuel and a new head-end process for fuel dissolution. The dissolution will be two-stage, using HF and HNO 3 , with an intermittent H 2 SO 4 dissolution for removing stainless steel components. Equipment operation is described

PWR fuel behavior: lessons learned from LOFT

International Nuclear Information System (INIS)

Russell, M.L.

1981-01-01

A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

Validating Westinghouse atom 16 x 16 and 18 x 18 PWR fuel performance

International Nuclear Information System (INIS)

Andersson, S.; Gustafson, J.; Jourdain, P.; Lindstroem, L.; Hallstadius, L.; Hofling, C.G.

2001-01-01

Westinghouse Atom designs and fabricates PWR fuel for all major European fuel types: 17 x 17 standard (12 ft) and 17 x 17 XL (14 ft) for Westinghouse type PWRs, and 16 x 16 and 18 x 18 fuel for Siemens type PWRs. The W Atom PWR fuel designs are based on the extensive Westinghouse CE PWR fuel experience from combustion engineering type PWRs. The W atom designs utilise basic design features from the W CE fuel tradition, such as all-Zircaloy mid grids and the proven ( 6 rod years) Guardian TM debris catcher, which is integrated in the bottom Inconel grid. Several new features have been developed to meet with stringent European requirements originating from requirements on very high burnup, in combination with low-leakage core operating strategies and high coolant temperatures. The overall reliability of the Westinghouse Atom PWR fuel is very high; no fuel failure has been detected since 1997. (orig.)

Modeling of PWR fuel at extended burnup

International Nuclear Information System (INIS)

Dias, Raphael Mejias

2016-01-01

This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

EDF/CIDEN - ONECTRA: PWR decontamination

International Nuclear Information System (INIS)

Fayolle, P.; Orcel, H.; Wertz, L.

2010-01-01

In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

Security of nuclear power in operation. Results from the first PWR 900 MWe stages of Electricity of France (EDF)

Energy Technology Data Exchange (ETDEWEB)

Capel, R; Chaubaron, J F [Electricite de France, 93 - Saint-Denis. Service de la Production Thermique

1980-06-01

The security and reliability objectives of the PWR 900 MWe stages are acquiring particular importance in the present energetic and nuclear context. This article presents the general framework wherein the superintendence and maintenance of plant equipment are situated. E.D.F. applies to all of its activities, the assurance of quality principles. The General Rules of Operating constitute the basic document. The Operating Technical Specifications specify the conditions for the correct operating and safety of the installations. The Organization of Quality handbook sets the rules to be obeyed in the management of all operations. Examples from Fessenhein and Bugey illustrate the subject and elucidate the practical dimension of security. Lastly, the lessons of experience are recalled.

Performance of high burned PWR fuel during transient

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Fujishiro, Toshio

1992-01-01

In a majority of Japanese light water type commercial powder reactors (LWRs), UO 2 pellet sheathed by zircaloy cladding is used. Licensed discharged burn-up of the PWR fuel rod is going to be increased from 39 MWd/kgU to 48 MWd/kgU. This requests the increased reliability of cladding material as a strong barrier against fission product (FP). A long time usage in the neutron field and in the high temperature coolant will cause the zircaloy hardening and embrittlement. The cladding material is also degraded by waterside corrosion. These degradations are enhanced much by increased burn-up. A increased magnitude of the pellet-cladding mechanical interaction (PCMI) is of importance for increasing the stress of cladding material. In addition, aggressive FPs released from the fuel tends to attack the cladding material to cause stress corrosion cracking (SCC). At the Nuclear Safety Research Reactor (NSRR) in JAERI, 14 x 14 PWR type fuel rods preirradiation up to 42 MWd/kgU was prepared for the transient pulse irradiation under the simulated reactivity initiated accident (RIA) conditions. This will cause a prompt increase of the fuel temperature and stress on the highly burned cladding material. In the present paper, steady-state and transient behavior observed from the tested PWR fuel rod and calculational results obtained from the computer code FPRETAIN will be described. (author)

The training for nuclear fuel handling at EDF

International Nuclear Information System (INIS)

Marion, J.P.

1999-01-01

The handling of fuel assemblies in a nuclear power plant presents 3 types of work: the taking delivery of fresh fuel, the refueling and the disposal of spent fuel. These operations are realized by teams made up of 3 handling operators and a supervisor. The refueling is made by 3*8-hour teams. These handling operations are important for the nuclear safety, a mishandling can damage the fuel cladding which is the first containment barrier, so a training center (CETIC) has been created. This center was founded in 1986 by EDF and Framatome, the purpose was to validate maintenance procedures, to test handling equipment and to train the teams which work on site. Various training programmes have been set up and a system of qualification degrees has been organized. The CETIC is fitted up with equipment that are full-sized mockups of real installations. Fuel assemblies don't react in a similar way to the different mechanical and neutronic stresses they undergo while they are in the core, they get deformed and the handling operations become more delicate. The mockup fuel assemblies are quite deformed to train the teams and prepare them to face any real situation. (A.C.)

The traveller: a new look for PWR fresh fuel packages

International Nuclear Information System (INIS)

Bayley, B.; Stilwell, W.E.; Kent, N.A.

2004-01-01

The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. This paper follows the development effort from the establishment of goals and objectives, to intermediate testing and analysis, to final testing and licensing. The discussion starts with concept origination and covers the myriad iterations that followed until arriving at a design that would meet the demanding licensing requirements, last for 30 years, and would be easy to load and unload fuel, easy to handle, inexpensive to manufacture and transport, and simple and inexpensive to maintain

Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

International Nuclear Information System (INIS)

Delafosse, Jacques.

1976-01-01

French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

Ruthenium release at high temperature from irradiated PWR fuels in various oxidising conditions. Main findings from the VERCORS program

International Nuclear Information System (INIS)

Ducros, G.; Pontillon, Y.; Malgouyres, P.P.; Taylor, P.; Dutheillet, Y.

2005-01-01

Fission product release and transport in case of PWR severe accident is a major topic in reactor safety assessment due to the potential radiological consequences for surrounding populations and the environment. In this context, the Institute for Radiological Protection and Safety (IRSN) and Electricite de France (EDF) have supported the VERCORS analytical test program which was performed by the ''Commissariat a l'Energie Atomique'' (CEA). It is usually considered as complementary to the PHEBUS FP in-pile integral experimental program. 25 annealing tests were performed between 1983 and 2002 on irradiated PWR fuels under various conditions of temperature and atmospheres (oxidising or reducing conditions).The influence of the nature of the fuel (UO 2 versus MOX, burn-up) and the fuel morphology (initially intact or fragmented fuels) have also been investigated. These led to an extended data base allowing on the one hand to study mechanisms which promote fission products release, and on the other hand to enhance models implemented in severe accident codes. Among all the fission products investigated, ruthenium is of specific concern because of its high radiological effects due essentially to the combination of both its short and long half-life isotopes (i.e. 103 Ru and 106 Ru respectively), but also by its ability to generate volatile gaseous oxides (RuO 3 , RuO 4 ) in very oxidising conditions, in particular in the case of air ingress accidents. Important uncertainties still remain on the release and transport of this element in such situations, and investigations on this open issue are notably carried out in the SARNET European framework. The present communication gives a general overview of the VERCORS program and presents more deeply the main findings concerning the ruthenium release. Its global behaviour is analysed on the basis of several comparative tests: same UO 2 sample (35 and 50 GWd/t) under hydrogen or steam conditions, similar MOX sample (40 GWd/t) under

EDF national emergency organization

International Nuclear Information System (INIS)

Laverge, J.

1992-01-01

The scope of French nuclear power led Electricite de France (EDF) to design standard products, manufactured in series but adaptable to different sites. Standardization is based on the decision on a single technology: pressurized water reactors (PWR). Thirty-four 900 MW and seventeen 1300 MW units are in operation on seventeen sites. The specific nature of French organization for normal operation and accident management results from equipment standardization and single licensee. This specificity is based on emergency plan standardization and highly structured national organization. Figs

Design and development of PWR fuel

International Nuclear Information System (INIS)

Dehon, C.; Leclercq, J.; Watteau, M.

1982-06-01

After a brief description of the FRAGEMA fuel assembly which equips at the present time the pressurized water reactors of EdF (Electricite de France), and a presentation of the experience obtained on this fuel, one reviews the main aims and trends of the research and development program carried out by FRAGEMA to improve the design of fuels and to propose to the national customer, but also on the foreign markets, new products adapted to the demands of operators. One insists more particularly on new products that are on one hand the AFA fuel and on the other hand the burnable poison UO 2 -Gd 2 O 3 ; their description is presented and their advantages are given. To conclude, one insists on the importance of the collaboration that have to be kept between the designer and the operator, the manufacturer, the R and D groups and the boiler specialist [fr

Evaluation model for PWR irradiated fuel

International Nuclear Information System (INIS)

Gomes, I.C.

1983-01-01

The individual economic value of the plutonium isotopes for the recycle of the PWR reactor is investigated, assuming the existence of an market for this element. Two distinct market situations for the stages of the fuel cycle are analysed: one for the 1972 costs and the other for costs of 1982. Comparisons are made for each of the two market situations concerning enrichment of the U-235 in the uranium fuel that gives the minimum cost in the fuel cycle. The method adopted to establish the individual value of the plutonium isotopes consists on the economical analyses of the plutonium fuel cycle for four different isotopes mixtures refering to the uranium fuel cycle. (Author) [pt

GAIA: AREVAs New PWR fuel assembly design

Energy Technology Data Exchange (ETDEWEB)

Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

2015-07-01

GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

Advancing PWR fuel to meet customer needs

Energy Technology Data Exchange (ETDEWEB)

Kramer, F W

1987-03-01

Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

International Nuclear Information System (INIS)

Fairclough, M.P.; Tymons, B.J.

1985-06-01

The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

The thermo-mechanics of the PWR fuel rod

International Nuclear Information System (INIS)

Barral, J.C.; Gautier, B.; Chaigne, G.

1999-01-01

The fuel rod mechanics is of a great importance in the safety and performance of the reactors. In this domain a meeting has been organized by the SFEN the 18 march 1998 at Paris. With the participation of scientists from CEA, EDF and Framatome, the physics of the fuel rods was presented based on four main aspects. Two first papers dealt with the solicitations of the fuel rod in normal and accidental conditions. The physical phenomena under irradiation were then detailed in the four following talks. Three papers presented the simulation and the codes of the fuel-cladding interactions with the diabolo effect. The last paper was devoted to the experiment feedback and the research programs. (A.L.B.)

EDF requirements for hot cells examinations on irradiated fuel

International Nuclear Information System (INIS)

Segura, J.C.; Ducros, G.

2002-01-01

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

Fuel operation of EDF nuclear fleet presentation of the centralized organization for operational engineering at the nuclear generation division

International Nuclear Information System (INIS)

Paulin, Ph.

2006-01-01

The main feature of EDF Nuclear Fleet is the standardization, with 'series' of homogeneous plants (same equipment, fuel and operation technical documents). For fuel operation, this standardization is related to the concept of 'fuel management scheme' (typical fuel reloads with fixed number and enrichment of fresh assemblies) for a whole series of plants. The context of the Nuclear Fleet lead to the choice of a centralized organization for fuel engineering at the Nuclear Generation Division (DPN), located at UNIPE (National Department for Fleet Operation Engineering) in Lyon. The main features of this organization are the following: - Centralization of the engineering activities for fuel operation support in the Fuel Branch of UNIPE, - Strong real-time link with the nuclear sites, - Relations with various EDF Departments in charge of design, nuclear fuel supply and electricity production optimization. The purposes of the organization are: - Standardization of operational engineering services and products, - Autonomy with independent methods and computing tools, - Reactivity with a technical assistance for sites (24 hours 'hot line'), - Identification of different levels (on site and off site) to solve core operation problems, - Collection, analysis and valorization of operation feedback, - Contribution to fuel competence global management inside EDF. This paper briefly describes the organization. The main figures of annual engineering production are provided. A selection of examples illustrates the contribution to the Nuclear Fleet performance. (authors)

Post irradiation examination on test fuel pins for PWR

International Nuclear Information System (INIS)

Fogaca Filho, N.; Ambrozio Filho, F.

1981-01-01

Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

Fuel reprocessing/fabrication interface

International Nuclear Information System (INIS)

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

1987-01-01

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

Power ramp testing method for PWR fuel rod at research reactor

International Nuclear Information System (INIS)

Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

2003-01-01

A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

The improvement of performances for PWR fuels

International Nuclear Information System (INIS)

Debes

2001-01-01

UO 2 fuels used in French nuclear power plants are authorized for values of burn-ups up to 52 GWj/t. Constant technological progress concerning pellets, cladding, and the design of the assembly has led to better performance and a broader safety margin. EDF is gathering all the elements to qualify and back its demand to increase the limit burn-up to 65 GWj/t in 2004 and to 70 GWj/t in 2008. For the same amount of energy produced, this policy of higher burn-ups will allow: - a reduction of the number of spent fuel assemblies, - a direct economic spare by using less fuel assemblies, - a reduction of personnel dosimetry because of longer irradiation campaigns, and - less quantity of residual plutonium produced. (A.C.)

EDF nuclear fuel cycle: report for the HCTISN meeting on the 20 of November 2009

International Nuclear Information System (INIS)

2009-01-01

The first part of this report describes the different steps of the EDF fuel cycle and the associated flows: overview, description of each step (natural uranium mines, enrichment, fabrication, production in reactors, processing, recycling). The second part deals with valuable materials and wastes (history of waste flows since 1994, MOX flows since 1994). It presents supply sources and flows (diversification to ensure supply safety, flows among different supply steps) and actions performed by EDF to promote information. Some propositions are given to improve information

Design of a PWR for long cycle and direct recycling of spent fuel

Energy Technology Data Exchange (ETDEWEB)

Mohamed, Nader M.A., E-mail: mnader73@yahoo.com

2015-12-15

Highlights: • Single-batch loading PWR with a new fuel assembly for 36 calendar months cycle was designed. • The new fuel assembly is constructed from a number of CANDU fuel bundles. • This design enables to recycle the spent fuel directly in CANDU reactors for high burnup. • Around 56 MWd/kgU burnup is achieved from fuel that has average enrichment of 4.8 w/o U-235 using this strategy. • Safety parameters such as the power distribution and CANDU coolant void reactivity were considered. - Abstract: In a previous work, a new design was proposed for the Pressurized Water Reactor (PWR) fuel assembly for direct use of the PWR spent fuel without processing. The proposed assembly has four zircaloy-4 tubes contains a number of 61-element CANDU fuel bundles (8 bundles per tube) stacked end to end. The space between the tubes contains 44 lower enriched UO{sub 2} fuel rods and 12 guide tubes. In this paper, this assembly is used to build a single batch loading 36-month PWR and the spent CANDU bundles are recycled in the on power refueling CANDU reactors. The Advanced PWR (APWR) is considered as a reference design. The average enrichment in the core is 4.76%w U-235. IFBA and Gd{sub 2}O{sub 3} as burnable poisons are used for controlling the excess reactivity and to flatten the power distribution. The calculations using MCNPX showed that the PWR will discharge the fuel with average burnup of 31.8 MWd/kgU after 1000 effective full power days. Assuming a 95 days plant outage, 36 calendar months can be achieved with a capacity factor of 91.3%. Good power distribution in the core is obtained during the cycle and the required critical boron concentration is less than 1750 ppm. Recycling of the discharged CANDU fuel bundles that represents 85% of the fuel in the assembly, in CANDU-6 or in 700 MWe Advanced CANDU Reactor (ACR-700), an additional burnup of about 31 or 26 MWd/kgU burnup can be achieved, respectively. Averaging the fuel burnup on the all fuel in the PWR

Preliminary study of the economics of enriching PWR fuel with a fusion hybrid reactor

International Nuclear Information System (INIS)

Kelly, J.L.

1978-09-01

This study is a comparison of the economics of enriching uranium oxide for pressurized water reactor (PWR) power plant fuel using a fusion hybrid reactor versus the present isotopic enrichment process. The conclusion is that privately owned hybrid fusion reactors, which simultaneously produce electrical power and enrich fuel, are competitive with the gaseous diffusion enrichment process if spent PWR fuel rods are reenriched without refabrication. Analysis of irradiation damage effects should be performed to determine if the fuel rod cladding can withstand the additional irradiation in the hybrid and second PWR power cycle. The cost competitiveness shown by this initial study clearly justifies further investigations

A comprehensive in-pile test of PWR fuel bundle

Energy Technology Data Exchange (ETDEWEB)

Kang Rixin; Zhang Shucheng; Chen Dianshan (Academia Sinica, Beijing (China). Inst. of Atomic Energy)

1991-02-01

An in-pile test of PWR fuel bundle has been conducted in HWRR at IAE of China. This paper describes the structure of the test bundle (3x3-2), fabrication process and quality control of the fuel rod, irradiation conditions and the main Post Irradiation Examination (PIE) results. The test fuel bundle was irradiated under the PWR operation and water chemistry conditions with an average linear power of 381 W/cm and reached an average burnup of 25010 MWd/tU of the fuel bundle. After the test, destructive and non-destructive examination of the fuel rods was conducted at hot laboratories. The fission gas release was 10.4-23%. The ridge height of cladding was 3 to 8 {mu}m. The hydrogen content of the cladding was 80 to 140 ppm. The fuel stack height was increased by 2.9 to 3.3 mm. The relative irradiation growth was about 0.11 to 0.17% of the fuel rod length. During the irradiation test, no fuel rod failure or other abnormal phenomena had been found by the on-line fuel failure monitoring system of the test loop and water sampling analysis. The structure of the test fuel assembly was left undamaged without twist and detectable deformation. (orig.).

Control in fabrication of PWR and BWR type reactor fuel elements

International Nuclear Information System (INIS)

Gorskij, V.V.

1981-01-01

Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

The development of flow test technology for PWR fuel assembly

International Nuclear Information System (INIS)

Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

1988-05-01

KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

Pulverulent deposits on fuel assemblies in primary circuit on 2 units at EDF (France)

International Nuclear Information System (INIS)

Zatla, A.; Piana, O.

2015-01-01

In march 2013, during the moving of some fuel assemblies in the fuel deactivation pool of two units nuclear power plants (Blayais 4 and Chinon B2 units), some fine and powdery deposits in suspension have been observed. In the two cases, the fuel has stayed in the deactivation pool during a long time (12 and 3 months), the units have started for the first time to inject Zinc in the primary circuit in the beginning of the precedent fuel cycle, and the units have operated an extended fuel cycle. EDF has performed analysis to evaluate the harmfulness of the particles and to investigate the mechanisms involved in this phenomenon. 3 conclusions can be drawn. First, the deposits are made up of usual corrosion products issued from the primary circuit components. Secondly, the chemical composition and the powdery characteristics of the deposits exclude nuclear safety risks. Thirdly, the atypical behaviour of the deposits could probably be linked with a modification of their structure due to a zinc effect. Because of the lack of new understanding elements, the extension of zinc injection to other NPP units has been broken off currently in France. EDF is studying the international experience feedback to identify the operating parameters of the nuclear power plant which might influent the phenomenon, and to evaluate if the deposits could enhance fuel cladding corrosion

MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

Energy Technology Data Exchange (ETDEWEB)

Herranz, L. E.; Lopez, C.

2013-07-01

The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

Quantitative analysis technique for Xenon in PWR spent fuel by using WDS

Energy Technology Data Exchange (ETDEWEB)

Kwon, H. M.; Kim, D. S.; Seo, H. S.; Ju, J. S.; Jang, J. N.; Yang, Y. S.; Park, S. D. [KAERI, Daejeon (Korea, Republic of)

2012-01-15

This study includes three processes. First, a peak centering of the X-ray line was performed after a diffraction for Xenon La1 line was installed. Xe La1 peak was identified by a PWR spent fuel sample. Second, standard intensities of Xe was obtained by interpolation of the La1 intensities from a series of elements on each side of xenon. And then Xe intensities across the radial direction of a PWR spent fuel sample were measured by WDS-SEM. Third, the electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to do matrix correction of a PWR spent fuel sample. Finally, the method and the procedure for local quantitative analysis of Xenon was developed in this study.

Quantitative analysis technique for Xenon in PWR spent fuel by using WDS

International Nuclear Information System (INIS)

Kwon, H. M.; Kim, D. S.; Seo, H. S.; Ju, J. S.; Jang, J. N.; Yang, Y. S.; Park, S. D.

2012-01-01

This study includes three processes. First, a peak centering of the X-ray line was performed after a diffraction for Xenon La1 line was installed. Xe La1 peak was identified by a PWR spent fuel sample. Second, standard intensities of Xe was obtained by interpolation of the La1 intensities from a series of elements on each side of xenon. And then Xe intensities across the radial direction of a PWR spent fuel sample were measured by WDS-SEM. Third, the electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to do matrix correction of a PWR spent fuel sample. Finally, the method and the procedure for local quantitative analysis of Xenon was developed in this study

Twenty-five years of transient counting experience in French PWR units

Energy Technology Data Exchange (ETDEWEB)

Barthelet, B. [Electricite de France (EDF DPN), 93 - Saint-Denis (France); Savoldelli, D.; Fritz, R. [Electricite de France (EDF DPN), 93 - Noisy le Grand (France)

2001-07-01

For nearly twenty five years, EDF has been checking that the actual operating transients are neither more severe nor more numerous than the design basis transients. This activity of transient cycle counting and bookkeeping has enabled EDF to own a database of more than 800 reactor.years for the PWR units. The current method of transient cycle counting is presented. In the paper, we will point out the main results of transient cycle counting and lessons learned. In general, the frequencies of transients are lower than the design frequencies. In few cases, they are higher, such as the transient frequencies of the RCS lines connected to auxiliary systems often due to operating procedures or particular periodic testing. Few periodic tests were not taken into account in the design basis transient file ; they have been detected thanks to the transient cycle counting. In the last 1980's, we achieved the first updating of the design basis transient file for the PWR 900 MWe series. In the early 1990's, we updated the design basis transient file of the PWR 1300 MWe series. In fact, since design and start-up, the operating conditions have been modified (fuel cycle with stretch-out, modification of the hot leg and cold leg temperatures for the PWR 1300 MWe,...). This was the cause of many unclassified transients. In the new design basis transient file, we have created new transients and increased the frequencies of some of them. This has enabled to consider the updated design basis transient file more representative of actual operating transients. For some years, we have increasingly associated the operators with the transient cycle counting concern. We noticed progress (decreased frequencies of most transients). (authors)

Twenty-five years of transient counting experience in French PWR units

International Nuclear Information System (INIS)

Barthelet, B.; Savoldelli, D.; Fritz, R.

2001-01-01

For nearly twenty five years, EDF has been checking that the actual operating transients are neither more severe nor more numerous than the design basis transients. This activity of transient cycle counting and bookkeeping has enabled EDF to own a database of more than 800 reactor.years for the PWR units. The current method of transient cycle counting is presented. In the paper, we will point out the main results of transient cycle counting and lessons learned. In general, the frequencies of transients are lower than the design frequencies. In few cases, they are higher, such as the transient frequencies of the RCS lines connected to auxiliary systems often due to operating procedures or particular periodic testing. Few periodic tests were not taken into account in the design basis transient file ; they have been detected thanks to the transient cycle counting. In the last 1980's, we achieved the first updating of the design basis transient file for the PWR 900 MWe series. In the early 1990's, we updated the design basis transient file of the PWR 1300 MWe series. In fact, since design and start-up, the operating conditions have been modified (fuel cycle with stretch-out, modification of the hot leg and cold leg temperatures for the PWR 1300 MWe,...). This was the cause of many unclassified transients. In the new design basis transient file, we have created new transients and increased the frequencies of some of them. This has enabled to consider the updated design basis transient file more representative of actual operating transients. For some years, we have increasingly associated the operators with the transient cycle counting concern. We noticed progress (decreased frequencies of most transients). (authors)

Layup practices at EDF PWR fleet and future developments

International Nuclear Information System (INIS)

Ramos, Olga Alos; Pablo, Sergio De Maria; Wintergerst, Matthieu

2014-01-01

A proper lay-up can only be achieved by plant specific lay-up procedures. The layup control practices are necessary to prevent corrosion damage during shutdown and are an important part of the life time extension of a nuclear power plant. EDF operational experience in this area confirms that damage to plant equipment from improper layup procedures is a cause of reduced plant reliability and availability and can increase the corrosion products ingress into the SG during start-up. The plant preservation strategy applied across the EDF fleet is based on the following principles: Anticipated time frame of the required layup: short-term outages can become long-term outages, and can necessitate the use of unplanned preservation techniques. Dry layup for a long-term outage of the condensate/feedwater system: condenser vacuum assisted draining at high temperature, estimation of air requirements (flow rate, humidity...) to guarantee the optimum conditions inside the system. Wet lay up for a short-term outage of the condensate/feedwater system: system filled with operating medium, avoiding air ingress into the system, injection of chemical agents if necessary. Dry lay-up or wet lay up for the steam generators depending on maintenance work or radiation protection needed. This article presents the EDF layup program and the improvements in layup practices. It provides information on the layup procedures, best practices and recommendations, operating experience, considerations for different durations of outages; the implementation of a preservation working group, the selection of the systems that will be included in the lay-up program. Finally, future developments will be presented, such as the development of a 'convective model' to determine air requirements (humidity, flow rate...) and the drying rate to dry a heat exchanger. (author)

EDF's Engineering Experience and Contribution to the Nuclear Development

International Nuclear Information System (INIS)

Salha, Bernard; Fourest, Bernard; Arpino, Jean-Marc

2002-01-01

Electricite de France (EDF) is now operating 58 nuclear power units which produce 76% of the electricity generated in France. This EDF's industrial success is the result of its capacity to master and optimize its production tool, from design through operation. EDF's integrated engineering is in the heart of this process of technical expertise and economic optimization. It allows to be in interface between the needs of operators and industrials suppliers, while accumulating a significant feedback of operating experience. The will of achieving the process of frenchifying PWR technology and to implement new industrial innovations have ended up in the new NPP of 100 % french design, the N4 series and its significant innovations. EDF energy policy is to keep the nuclear option open for the future. This strategy results from the need to improve the availability and the life extension of the units in operation and to prepare the replacement of the operating reactors around 2015. This is the objective of the European Pressurized Reactor (EPR), a French-German joint project. EDF is also applying this industrial process in its international projects. For example China, which desires to implement a standardized nuclear program and to move forward the complete autonomy of its nuclear industry, has decided to adopt a similar approach to EDF's one. (authors)

The optimum fuel and power distribution for a PWR burnup cycle

International Nuclear Information System (INIS)

Stillman, J.A.

1989-01-01

A method was developed to determine the optimum fuel and power distributions for a PWR burnup cycle. The backward diffusion calculation [1] and the Core-wise Green's Function [2] method were used for the core model which provided analytic derivatives for solving the nonlinear optimization problem using successive linear programming [3] methods. The solution algorithm consisted of a reverse depletion strategy which begins at the end of cycle and solves simultaneously for the optimal fuel and burnable absorber distributions while the core is depleted to the beginning of cycle. The resulting optimal solutions minimize the required fissile fuel inventory and burnable absorber loading for a PWR

Development and testing of the EDF-2 reactor fuel element; Essais et mise au point de l'element combustible pour le reacteur EDF-2

Energy Technology Data Exchange (ETDEWEB)

Delpeyroux, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Furhmann, R [Societe Industrielle de Combustible Nucleaire (France)

1964-07-01

This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the {beta} phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [French] Ce rapport technique

Safety experience on EDF's PWRs

International Nuclear Information System (INIS)

Tanguy, P.

1986-01-01

The french nuclear programme has been widely publicized. In 1985, the total nuclear electricity generated was around 216 GWh, i. e. 70% of the electricity produced by electricity de France (EDF). If we consider only pressurized water reactors, at the end of 1985, 37 units were in operation (32 900 MWe and 5 1300 MWe) and 18 were under construction. I intend to review our experience with the safety of PWR's, but I will first present briefly some aspects related to the safety organization in France and the standardization policy. (author) [pt

Benchmarking Computational Fluid Dynamics for Application to PWR Fuel

International Nuclear Information System (INIS)

Smith, L.D. III; Conner, M.E.; Liu, B.; Dzodzo, B.; Paramonov, D.V.; Beasley, D.E.; Langford, H.M.; Holloway, M.V.

2002-01-01

The present study demonstrates a process used to develop confidence in Computational Fluid Dynamics (CFD) as a tool to investigate flow and temperature distributions in a PWR fuel bundle. The velocity and temperature fields produced by a mixing spacer grid of a PWR fuel assembly are quite complex. Before using CFD to evaluate these flow fields, a rigorous benchmarking effort should be performed to ensure that reasonable results are obtained. Westinghouse has developed a method to quantitatively benchmark CFD tools against data at conditions representative of the PWR. Several measurements in a 5 x 5 rod bundle were performed. Lateral flow-field testing employed visualization techniques and Particle Image Velocimetry (PIV). Heat transfer testing involved measurements of the single-phase heat transfer coefficient downstream of the spacer grid. These test results were used to compare with CFD predictions. Among the parameters optimized in the CFD models based on this comparison with data include computational mesh, turbulence model, and boundary conditions. As an outcome of this effort, a methodology was developed for CFD modeling that provides confidence in the numerical results. (authors)

EDF program for cleaning of PWR high dosimetric NPPs results and perspectives of radiological cleanup engineering

International Nuclear Information System (INIS)

Ider, Samir

2012-09-01

In 2002 EDF's PWRs presented significant disparities in their integrated dose results. In order to treat the PWR with highest dose per shutdown ratio, EDF/DPN (Nuclear Production Division) decided to develop an engineering project for 'PWR high dosimetry cleaning'. A specific financial support and staff resource was allocated to each reactor on this project. On every reactor, equipment is decontaminated using chemical process. The average dose reduction is estimated at 0,5 Man.mSv per reactor treated over the five following years. During the power operation of a nuclear reactor, corrosion products from the primary circuit undergo activation through their circulation in the reactor core. These activated products can be integrated into the oxide layer that builds up during the power operation on the primary piping. Additionally, highly radioactive particles can fall by gravity to the bottom of the reactor vessel. Being highly mobile, these can be dispersed throughout the circuit system by the water movement during the shutdown of the Unit. These particles will become hot spots when trapped in specific features of the primary cooling circuit and its auxiliaries. This contamination of the primary circuit leads to high dosimetry of the personnel during the maintenance of the Unit. To decrease the dose received during the maintenance operations, curative actions must be taken to eliminate hot spots and reduce contamination, as well as preventatives measures to curb future new contamination. For each PWR concerned, the studies begin with a diagnostic of the origins of the contamination, the type of pollution, its form (mobile hot spots, oxide layers) and its location. Many types of measuring equipment are used to make a radiological diagnostic (portable gamma-ray spectroscopy, video cartography, and gamma camera system). Experience and knowledge of the underlying chemistry in decontamination process, leads to a Decontamination Factor (DF) per system for

New package systems for EDF recycling policy

International Nuclear Information System (INIS)

Leboucher, I.; Michels, L.; Hugon, F.C.; Penoty, C.; Froment, A.

2004-01-01

Since the beginning of the nineties, EDF (Electricite France) has been carrying out a policy towards increasing burn-up. As a consequence and to keep energetic performances of the MOX (Mixed OXide uranium and plutonium) fresh fuel assemblies at parity with UO2 fuel, the concentration in plutonium increased up to 7.1%, soon up to 8.65% and even more. To take into account the increased dose rate and the increased thermal power of these MOX fresh fuels in the transportation system, COGEMA LOGISTICS designed, manufactured and licensed the MX8 package to replace the previous packaging. Simultaneously new high-secured transport system was developed. Due to the capacity of the reactor pool, the spent fuel assemblies must be evacuated with a short cooling time and high residual power. Because of the neutron source of these future MOX spent fuel assemblies, the TN 12/2 packaging, which has been in operation since 1980's to transport UO2 or MOX spent fuel to COGEMA La Hague, from EDF, other European and Japanese NPP lacks flexibility. COGEMA LOGISTICS is designing a new cask (TN trademark 112) to transport the future EDF MOX spent fuel assemblies. The TN trademark 112 has to comply with the 1996 edition of the IAEA transport regulations and with the transport and handling equipment presently used for the transport of spent fuel in France. It can be loaded with 12 assemblies, either MOX or UO2. This paper gives a description of two packages belonging to the new generation: - one for fresh MOX fuel assemblies transportation, used on EDF's power plants for MOX reloads, - one for spent MOX fuel assemblies transportation, presently under development with a first transport planned for mid-2007

New package systems for EDF recycling policy

Energy Technology Data Exchange (ETDEWEB)

Leboucher, I.; Michels, L.; Hugon, F.C. [COGEMA Logistics (AREVA Group) (France); Penoty, C. [EDF - Electricite de France (France); Froment, A. [COGEMA (AREVA Group) (France)

2004-07-01

Since the beginning of the nineties, EDF (Electricite France) has been carrying out a policy towards increasing burn-up. As a consequence and to keep energetic performances of the MOX (Mixed OXide uranium and plutonium) fresh fuel assemblies at parity with UO2 fuel, the concentration in plutonium increased up to 7.1%, soon up to 8.65% and even more. To take into account the increased dose rate and the increased thermal power of these MOX fresh fuels in the transportation system, COGEMA LOGISTICS designed, manufactured and licensed the MX8 package to replace the previous packaging. Simultaneously new high-secured transport system was developed. Due to the capacity of the reactor pool, the spent fuel assemblies must be evacuated with a short cooling time and high residual power. Because of the neutron source of these future MOX spent fuel assemblies, the TNa12/2 packaging, which has been in operation since 1980's to transport UO2 or MOX spent fuel to COGEMA La Hague, from EDF, other European and Japanese NPP lacks flexibility. COGEMA LOGISTICS is designing a new cask (TN trademark 112) to transport the future EDF MOX spent fuel assemblies. The TN trademark 112 has to comply with the 1996 edition of the IAEA transport regulations and with the transport and handling equipment presently used for the transport of spent fuel in France. It can be loaded with 12 assemblies, either MOX or UO2. This paper gives a description of two packages belonging to the new generation: - one for fresh MOX fuel assemblies transportation, used on EDF's power plants for MOX reloads, - one for spent MOX fuel assemblies transportation, presently under development with a first transport planned for mid-2007.

Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios

International Nuclear Information System (INIS)

Zhou, Shengcheng; Wu, Hongchun; Zheng, Youqi

2018-01-01

Highlights: •ADS reloading scheme is optimized to raise discharge burnup and lower reactivity loss. •ADS is flexible to be combined with various pyro-chemical reprocessing technologies. •ADS is flexible to transmute MAs from different spent PWR fuels. -- Abstract: A two-stage Pressurized Water Reactor (PWR)-Accelerator Driven System (ADS) fuel cycle is proposed as an option to transmute minor actinides (MAs) recovered from the spent PWR fuels in the ADS system. At the second stage, the spent fuels discharged from ADS are reprocessed by the pyro-chemical process and the recovered actinides are mixed with the top-up MAs recovered from the spent PWR fuels to fabricate the new fuels used in ADS. In order to lower the amount of nuclear wastes sent to the geological repository, an optimized scattered reloading scheme for ADS is proposed to maximize the discharge burnup and lower the burnup reactivity loss. Then the flexibility of ADS for MA transmutation is evaluated in this research. Three aspects are discussed, including: different cooling time of spent ADS fuels before reprocessing, different reprocessing loss of spent ADS fuels, and different top-up MAs recovered from different kinds of spent PWR fuels. The ADS system is flexible to be combined with various pyro-chemical reprocessing technologies with specific spent fuels cooling time and unique reprocessing loss. The reduction magnitudes of the long-term decay heat and radiotoxicity of MAs by transmutation depend on the reprocessing loss. The ADS system is flexible to transmute MAs recovered from different kinds of spent PWR fuels, regardless of UOX or MOX fuels. The reduction magnitudes of the long-term decay heat and radiotoxicity of different MAs by transmutation stay on the same order.

Transuranics and fission products release from PWR fuels in severe accident conditions. Lessons learnt from VERCORS RT3 and RT4 tests

International Nuclear Information System (INIS)

Pontillon, Y.; Ducros, G.; Van Winckel, S.; Christiansen, B.; Kissane, M.P.; Dubourg, R.; Dutheillet, Y.; Andreo, F.

2006-01-01

Over the last decades, several experimental programs devoted to the source term of fission products (FP) and actinides released from PWR fuel samples in severe accident (SA) conditions have been initiated throughout the world. In France, in this context, the Institute for Radiological Protection and Safety (IRSN) and Electricite de France (EDF) have supported the analytical VERCORS program which was performed by the Commissariat a l'Energie Atomique (CEA). The VERCORS facility at the LAMA-laboratory (CEA-Grenoble, France) was designed to heat up an irradiated fuel sample - taken from EDF's nuclear power reactors - to fuel relocation, and to capture the fission products released from the fuel and deposited downstream on a series of specific filters (impactors, bead-bed filter). On-line gamma detectors aimed at the fuel position, filters and gas capacity monitored the progress of FP release from the fuel, FP deposition on the filters and the fission gases emitted by the fuel (xenon and krypton). Before and after the test, a longitudinal gamma-scan of the fuel was conducted to measure the initial and final FP inventory in order to evaluate the quantitative fractions of FP emitted by the fuel during the test. All the components of the loop were then gamma-scanned to measure and locate the FPs released during the test and to draw up a mass balance of these FP. 25 annealing tests were performed between 1983 and 2002 on irradiated PWR fuels under various conditions of temperature and atmospheres (oxidising or reducing conditions). The influence of the nature of the fuel (UO 2 versus MOX, burn up) and the fuel morphology (initially intact or fragmented fuel) have also been investigated. This led to an extended data base allowing on the one hand to study mechanisms which promote FP release in SA conditions, and on the other hand to enhance models implemented in SA codes. Because gamma spectrometry is well suited to FP measurement and not to actinides (except neptunium

Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

International Nuclear Information System (INIS)

Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

2015-01-01

Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

Energy Technology Data Exchange (ETDEWEB)

Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

2015-12-15

A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

Axial gap formation in P.W.R. fuel pins

International Nuclear Information System (INIS)

Roberts, G.; Jones, K.W.

1978-07-01

The potential mechanisms of axial gap formation in PWR fuel pins are examined analytically and also using evidence from post-irradiation examination (p.i.e.) investigation. It is concluded that fuel and cladding cannot remain in contact during densification and so the settling of of the fuel stack, which forms the gaps, must be prevented by such things as asperities in the cladding, fuel chips or tilted pellets. Examples from the p.i.e. examination programme are used to support this conclusion. (author)

Fire experiences: principal lessons learned, application in PWR power plants

International Nuclear Information System (INIS)

Schoemacker, M.

1984-01-01

The article reviews the principal design rules to be borne in mind for PWR nuclear units installation. These rule takes into account: the specific character of materials involved (safety aspect for nuclear construction), experience acquired as a result of fires in EDF production units, and the results obtained from tests carried out by the EDF at Fort de Chelles between 1980 and 1982, especially in the field of PVC cables [fr

Analysis of reactivity worths of highly-burnt PWR fuel samples measured in LWR-PROTEUS Phase II

Energy Technology Data Exchange (ETDEWEB)

Grimm, Peter; Murphy, Michael F.; Jatuff, Fabian; Seiler, Rudolf [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland)

2008-07-01

The reactivity loss of PWR fuel with burnup has been determined experimentally by inserting fresh and highly-burnt fuel samples in a PWR test lattice in the framework of the LWR-PROTEUS Phase II programme. Seven UO{sub 2} samples irradiated in a Swiss PWR plant with burnups ranging from approx40 to approx120 MWd/kg and four MOX samples with burnups up to approx70 MWd/kg were oscillated in a test region constituted of actual PWR UO{sub 2} fuel rods in the centre of the PROTEUS zero-power experimental facility. The measurements were analyzed using the CASMO-4E fuel assembly code and a cross section library based on the ENDF/B-VI evaluation. The results show close proximity between calculated and measured reactivity effects and no trend for a deterioration of the quality of the prediction at high burnup. The analysis thus demonstrates the high accuracy of the calculation of the reactivity of highly-burnt fuel. (authors)

PWR life time: the EDF project

International Nuclear Information System (INIS)

Noel, R.; Reynes, L.; Mercier, J.P.

1987-01-01

Operating a very large number of standardized PWR units which supply today 70% of French power generation, Electricite de France is highly interested in getting the best estimate of the safe and economical life of these plants. An extensive program of work has been undertaken in this respect. The studies have first to go through all available data on aging process, survey and maintenance of a limited number of major components. This review will lead to recommendation of complementary work in these fields. The first conclusions are that these units are able to perform a long service time, under provision of careful survey and maintenance [fr

Material control and accountancy at EDF PWR plants

International Nuclear Information System (INIS)

de Cormis, F.

1991-01-01

The paper describes the comprehensive system which is developed and implemented at Electricite de France to provide a single reliable nuclear material control and accounting system for all nuclear plants. This software aims at several objectives among which are: the control and the accountancy of nuclear material at the plant, the optimization of the consistency of data by minimizing the possibility of transcription errors, the fulfillment of the statutory requirements by automatic transfer of reports to national and international safeguards authorities, the servicing of other EDF users of nuclear material data for technical or commercial purposes

Modeling of the thermo-mechanical behaviour of the PWR fuel

International Nuclear Information System (INIS)

Mailhe, P.

2014-01-01

This article reviews the various physical phenomena that take place in an irradiated fuel rod and presents the development of the thermo-mechanical codes able to simulate them. Though technically simple the fuel rod is the place where appear 4 types of process: thermal, gas behaviour, mechanical and corrosion that combine involving 5 elements: the fuel pellet, the fuel clad, the fuel-clad gap, the inside volume and the coolant. For instance the pellet is the place where the following mechanical processes took place: thermal dilatation, elastic deformation, creep deformation, densification, solid swelling, gaseous swelling and cracking. The first industrial code simulating the behaviour of the fuel rod was COCCINEL, it was developed by AREVA teams from the American PAD code that was included in the Westinghouse license. Today the GALILEO code has replaced the COPERNIC code that was developed in the beginning of the 2000 years. GALILEO is a synthesis of the state of the art of the different models used in the codes validated for PWR and BWR. GALILEO has been validated on more than 1500 fuel rods concerning PWR, BWR and specific reactors like Siloe, Osiris, HFR, Halden, Studsvik, BR2/3,...) and also for extended burn-ups. (A.C.)

A comparative study of fuel management in PWR reactors

International Nuclear Information System (INIS)

Barroso, D.E.G.; Nair, R.P.K.; Vellozo, S.O.

1981-01-01

A study about fuel management in PWR reactors, where not only the conventional uranium cycle is considered, but also the thorium cycle as an alternative is presented. The final results are presented in terms of U 3 O 8 demand and SWU and the approximate costs of the principal stages of the fuel cycle, comparing with the stardand cycle without recycling. (E.G.) [pt

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods

International Nuclear Information System (INIS)

Schienbein, Marcel; Zeh, Peter; Hurtado, Antonio; Rosskamp, Matthias; Mailand, Irene; Bolz, Michael

2012-09-01

In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing

Optimum fuel use in PWR reactors

International Nuclear Information System (INIS)

Neubauer, W.

1979-07-01

An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.) [de

The effects of radiations on materials for core internals of PWRs: EDF-CEA-Framatome joint research programme

International Nuclear Information System (INIS)

Mathan, N. de; Buisine, D.; Goltrant, O.; Dubuisson, P.; Scott, P.; Deydier, D.; Trenty, A.

1998-01-01

The effects of neutron irradiation on materials for the core internals of PWRs (austenitic stainless steels) are potentially a significant economic and regulatory concern for EDF. The maintenance strategy for EDF relies primarily on in-service inspection, safety analysis and characterization of materials irradiated in-service. In addition, to anticipate likely future behaviour of highly irradiated materials, EDF has initiated, in collaboration with CEA and Framatome, a large R and D programme designed to (i) evaluate the effects of neutron irradiation on mechanical properties and stress corrosion cracking sensitivity (IASCC), and (ii) identify possible replacement materials. The programme, currently in progress, involves mechanical tests (tensile, fracture toughness, irradiation creep), stress corrosion cracking tests (in flux and out of flux) and metallurgical examinations. The test materials are being irradiated in several experimental reactors in France and Russia up to PWR-related end of life doses (∼ 80 dpa) at several PWR-relevant irradiation temperatures (300-400 deg. C). The presentation will describe the objectives and early results of this ongoing R and D programme. (author)

Lay-up practices at EDF PWR fleet and future developments

International Nuclear Information System (INIS)

Alos Ramos, O.; Wintergest, M.; De Maria Pablo, S.; Postic, A.; Courtaudiere, J.P.; Fournier, C.; Vialette, R.; Waeber, J.; Couturier, D.; Robillart, P.

2015-01-01

A proper lay-up can only be achieved by plant specific lay-up procedures. The lay-up control practices are necessary to prevent corrosion damage during shut down and are an important part of the life extension of a nuclear power plant. EDF operational experience in this area confirms that damage to plant equipment from improper lay-up procedures is a cause of reduced plant reliability and availability and can increase the corrosion products ingress into the SG during start-up. This article presents the EDF lay-up program and the improvement of lay-up practices. It provides information on the lay-up procedures, best practices and recommendations, operating experience, considerations for different durations of outages; the implementation of a preservation working group, the selection of the systems that will be included in the lay-up program and the material damage characterization in some auxiliary systems. Finally, future developments will be presented, such as the development of a 'convective model' to determine air requirements (humidity, flow rate...) and the drying rate to dry a heat exchanger. (authors)

Analysis of difficulties accounting and evaluating nuclear material of PWR fuel plant

International Nuclear Information System (INIS)

Zhang Min; Jue Ji; Liu Tianshu

2013-01-01

Background: Nuclear materials accountancy must be developed for nuclear facilities, which is required by regulatory in China. Currently, there are some unresolved problems for nuclear materials accountancy of bulk nuclear facilities. Purpose: The retention values and measurement errors are analyzed in nuclear materials accountancy of Power Water Reactor (PWR) fuel plant to meet the regulatory requirements. Methods: On the basis of nuclear material accounting and evaluation data of PWR fuel plant, a deep analysis research including ratio among random error variance, long-term systematic error variance, short-term systematic error variance and total error involving Material Unaccounted For (MUF) evaluation is developed by the retention value measure in equipment and pipeline. Results: In the equipment pipeline, the holdup estimation error and its total proportion are not more than 5% and 1.5%, respectively. And the holdup estimation can be regraded as a constant in the PWR nuclear material accountancy. Random error variance, long-term systematic error variance, short-term systematic error variance of overall measurement, and analytical and sampling methods are also obtained. A valuable reference is provided for nuclear material accountancy. Conclusion: In nuclear material accountancy, the retention value can be considered as a constant. The long-term systematic error is a main factor in all errors, especially in overall measurement error and sampling error: The long-term systematic errors of overall measurement and sampling are considered important in the PWR nuclear material accountancy. The proposals and measures are applied to the nuclear materials accountancy of PWR fuel plant, and the capacity of nuclear materials accountancy is improved. (authors)

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)

Comparison of PWR-IMF and FR fuel cycles

International Nuclear Information System (INIS)

Darilek, Petr; Zajac, Radoslav; Breza, Juraj; Necas, Vladimir

2007-01-01

The paper gives a comparison of PWR (Russia origin VVER-440) cycle with improved micro-heterogeneous inert matrix fuel assemblies and FR cycle. Micro-heterogeneous combined assembly contains transmutation pins with Pu and MAs from burned uranium reprocessing and standard uranium pins. Cycle analyses were performed by HELIOS spectral code and SCALE code system. Comparison is based on fuel cycle indicators, used in the project RED-IMPACT - part of EU FP6. Advantages of both closed cycles are pointed out. (authors)

Application of burnup credit for PWR spent fuel storage pool

International Nuclear Information System (INIS)

Shin, Hee Sung; Ro, Seung-Gy; Bae, Kang Mok; Kim, Ik Soo; Shin, Young Joon

1999-01-01

A study on the application of burnup credit for a PWR spent fuel storage pool has been investigated using a computer code system such as CSAS6 module of SCALE 4.3 in association with 44-group SCALE cross-section library. The calculation bias of the code system at a 95% probability with a 95% confidence level seems to be 0.00951 by benchmarking the system for forty six experimental data. With the aid of this computer code system, criticality analysis has been performed for the PWR spent fuel storage pool. Uncertainties due to postulated abnormal and accidental conditions, and manufacturing tolerance such as stainless steel thickness of storage rack, fuel enrichment, fuel density and box size have statistically been combined and resulted in 0.00674. Also, isotopic correction factor which was based on the calculated and measured concentration of 43 isotopes for both selected actinides and fission products important in burnup credit application has been taken into account in the criticality analysis. It is revealed that the minimum burnup with the corrected isotopic concentrations as required for the safe storage is 5,730 MWd/tU in enriched fuel of 5.0 wt%. (author)

Generation of SCALE 6 Input Data File for Cross Section Library of PWR Spent Fuel

International Nuclear Information System (INIS)

Jeong, Chang Joon; Cho, Dong Keun

2010-11-01

In order to obtain the cross section libraries of the Korean Pressurized water reactor (PWR) spent fuel (SF), SCALE 6 code input files have been generated. The PWR fuel data were obtained from the nuclear design report (NDR) of the current operating PWRs. The input file were prepared for 16 fuel types such as 4 types of Westinghouse 14x14, 3 types of OPR-1000 16x16, 4 types of Westinghouse 16x16, and 6 types of Westinghouse 17x17. For each fuel type, 5 kinds of fuel enrichments have been considered such as 1.5, 2.0 ,3.0, 4.0 and 5.0 wt%. In the SCALE 6 calculation, a ENDF-V 44 group was used. The 25 burnup step until 72000 MWD/T was used. A 1/4 symmetry model was used for 16x16 and 17x17 fuel assembly, and 1/2 symmetry model was used for 14x14 fuel assembly The generated cross section libraries will be used for the source-term analysis of the PWR SF

Ciclon: A neutronic fuel management program for PWR's consecutive cycles

International Nuclear Information System (INIS)

Aragones, J.M.

1977-01-01

The program description and user's manual of a new computer code is given. Ciclon performs the neutronic calculation of consecutive reload cycles for PWR's fuel management optimization. Fuel characteristics and burnup data, region or batch sizes, loading schemes and state of previously irradiated fuel are input to the code. Cycle lengths or feed enrichments and burnup sharing for each region or batch are calculate using different core neutronic models and printed or punched in standard fuel management format. (author) [es

PWR fuel inspection and repair technology development in the Republic of Korea

International Nuclear Information System (INIS)

Park, J.Y.

1998-01-01

As of September 1997, 10 PWRs and 2 PHWRs generate 10,320MW electricity in Korea. And another 8 PWRs and 2 PHWRs will be constructed by 2006. These will need about 400 MTU of PWR fuels and 400 MTU of PHWR fuels. To improve average burnup, thermal power, fuel usability and plant safety, better poolside fuel service technologies are strongly recommended as well as the fuel design and fabrication technology improvements. During the last twenty years of nuclear power plant operation in Korea, more than 4,000 fuel assemblies has been used. At the site, continuous coolant activity measurement, pool-side visual inspection and ultrasonic tests have been performed. Some of the fuels are damaged or failed for various reasons. Some of the defected fuels were examined in hot cell to investigate the cause of failure. Even though 30 PWR fuel assemblies were repaired by foreign engineers, fuel inspection and repair technologies are not established yet. Various kind of design for the fuel make the inspection, repair and reconstitution equipment more complex. As a result, recently, a plant to obtain overall technology for poolside fuel inspection, failed fuel repair and reconstitution through R and D activities are set forth. (author)

Ultrasonic inspection for testing the PWR fuel rod endplug welds

International Nuclear Information System (INIS)

Pillet, C.; Destribats, M.T.; Papezyk, F.

1976-01-01

A method of ultrasonic testing with local immersion and transversal waves was developed. It is possible to detect defects as the lacks of fusion and penetration and porosity in the PWR fuel rod endplug welds [fr

Station blackout: Deterministic and probabilistic approach in the field of electrical supply losses by EDF

International Nuclear Information System (INIS)

Meslin, T.; Carnino, A.

1986-01-01

This example shows the thoroughness of EDF's approach in processing the difficult problems of the loss of electrical power supplies. Efforts are continuing in several directions: continued revision and improvement of operating procedures in the event of loss of electrical power supplies, PWR plant operator training courses devoted to the problems of power supply losses, and continued testing on simulators, and particularly testing under real conditions, including tests lasting several hours made possible by the performance of the new EDF simulators (two-phase code and taking all power losses into account)

Thermal analysis of a one-element PWR spent fuel shipping cask

International Nuclear Information System (INIS)

Fields, S.R.

1979-06-01

The transient thermal behavior of a typical one-element PWR spent fuel shipping cask, following a hypothetical accident and fire, has been simulated. The objectives of the study were to determine the transient behavior of the cask and its spent fuel primary coolant through the pressure relief system and possible fuel pin clad failure due to overheating following loss of coolant. 15 figures, 7 tables

Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

Energy Technology Data Exchange (ETDEWEB)

Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N. [Westinghouse Electric Co. LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existing facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)

A new uncertainty reduction method for PWR cores with erbia bearing fuel

International Nuclear Information System (INIS)

Takeda, Toshikazu; Sano, Tadafumi; Kitada, Takanori; Kuroishi, Takeshi; Yamasaki, Masatoshi; Unesaki, Hironobu

2008-01-01

The concept of a PWR with erbia bearing high burnup fuel has been proposed. The erbia is added to all fuel with over 5% 235 U enrichment to retain the neutronics characteristics to that within 5% 235 U enrichment. There is a problem of the prediction accuracy of the neutronics characteristics with erbia bearing fuel because of the short of experimental data of erbia bearing fuel. The purpose of the present work is to reduce the uncertainty. A new method has been proposed by combining the bias factor method and the cross section adjustment method. For the PWR core, the uncertainty reduction, which shows the rate of reduction of uncertainty, of the k eff is 0.865 by the present method and 0.801 by the conventional bias factor method. Thus the prediction uncertainties are reduced by the present method compared to the bias factor method. (authors)

Zinc injection implementation process at EDF: risk analysis, chemical specifications and operating procedures

International Nuclear Information System (INIS)

Tigeras, A.; Stutzmann, A.; Bremnes, O.; Claeys, M.; Ranchoux, G.; Segura, J.C.; Errera, J.; Bonne, S.

2010-01-01

Zinc's ability to replace cobalt from oxides of primary circuit surfaces has provided the first motivation for implementing the zinc addition in BWRs since the mid-1980s. The beneficial results regarding dose reductions have been demonstrated; therefore, this practice has been extended to PWRs since the 1990s, not only for radiation fields considerations, but also for reducing PWSCC. From the beginning of the 2000s, further reasons to inject zinc associated with the fuel management process have been identified (e; g; power increase, high burn-up, and/or cycle length increase). These evolutions must be accompanied by an adapted chemistry program in order to mitigate the crud deposition on fuel assemblies and the consequent AOA/CIPS or localized clad corrosion risks. The source term reduction (due to the decrease of the general corrosion rate of several materials) and the absence of negative impact on alloy cladding in the presence of zinc in the primary coolant are the main reasons for selecting zinc injection as a reliable option for preventing and/or mitigating the effects of fuel deposits. These three PWR motivations (field radiation, components performance, and fuel reliability) are also the major objectives of CANDU®, WWER, and new reactors (EPR, AP1000), where the zinc injection feasibility analyses are in progress in order to improve the safety of their operating conditions. With the purpose of achieving the optimal results of zinc injection, the process's implantation in a unit must be conducted with an appropriate risk analysis, covering all possible domains affected by this primary coolant chemistry modification : safety, fuel and component performance, radioprotection, waste, environment, human and installation security, human and material resources, staff formation, and documentation. EDF has performed a complete analysis of this enlarged scope, relying upon theoretical and experimental results as well as NPP feedback. This paper describes EDF

Development of MHI PWR fuel assembly with high thermal performance

International Nuclear Information System (INIS)

Yasushi Makino; Masaya Hoshi; Masaji Mori; Hidetoshi Kido; Kazuo Ikeda

2005-01-01

Mitsubishi Heavy Industries, Ltd. (MHI) has been developing a PWR fuel assembly to meet the needs of Japanese fuel market with mainly improving its reliability such as a mechanical strength, a seismic strength and endurance. For burn-up extension of the fuel to 55 GWd/t, MHI has introduced a Zircaloy spacer grid with better neutron economics with retaining the reliability in an operating core. However, for a future power up-rating and a longer cycle operation, a higher thermal performance is required for PWR fuel assembly. To meet the needs of fuel market, MHI has developed an advanced type of Zircaloy spacer grid with a greater DNB performance while retaining the reliability of a fuel and a relatively low pressure drop. For the greater DNB performance, MHI optimized geometrical shape of mixing vane to promote a fluid mixing performance. In this report, higher DNB performance provided by the advanced Zircaloy spacer grid is presented. The results of 3D simulation for the flow behavior in 5 x 5 partial assembly, a mixing test and a water DNB test were compared between the current and the advanced spacer grids. Consequently, it was confirmed that a crossover vane enhanced a fluid mixing and the advanced spacer grid could significantly improve DNB performance compared with the current design of spacer grids. (authors)

Uranium savings on a once through PWR fuel cycle

International Nuclear Information System (INIS)

Cupo, J.V.

1980-01-01

A number of alternatives which have the greatest potential for near term savings with minimum plant and fuel modifications have been examined at Westinghouse as part of continued internal assessment and part of NASAP study conducted for DOE pertaining to uranium utilization in a once through PWR fuel cycle. The alternatives which could be retrofitted to existing reactors were examined in more detail in the evaluation since they would have the greater near term impact on U savings

Fuel rod behavior of a PWR during load following

International Nuclear Information System (INIS)

Perrotta, J.A.; Andrade, G.G. de

1982-01-01

The behavior of a PWR fuel rod when operating in normal power cycles, excluding in case of accidents, is analysed. A computer code, that makes the mechanical analysis of the cladding using the finite element method was developed. The ramps and power cycles were simulated suposing the existence of cracks in pellets when the cladding-pellet interaction are done. As a result, an operation procedure of the fuel rod in power cycle is recommended. (E.G.) [pt

ABB PWR fuel design for high burnup

International Nuclear Information System (INIS)

Nilsson, S.; Jourdain, P.; Limback, M.; Garde, A.M.

1998-01-01

Corrosion, hydriding and irradiation induced growth of a based materials are important factors for the high burnup performance of PWR fuel. ABB has developed a number of Zr based alloys to meet the need for fuel that enables operation to elevated burnups. The materials include composition and processing optimised Zircaloy 4 (OPTIN TM ) and Zircaloy 2 (Zircaloy 2P), as well as advanced Zr based alloys with chemical compositions outside the composition specified for Zircaloy. The advanced alloys are either used as Duplex or as single component claddings. The Duplex claddings have an inner component of Zircaloy and an outer layer of Zr with small additions of alloying elements. ABB has furthermore improved the dimensional stability of the fuel assembly by developing stiffer and more bow resistant guide tubes while debris related fuel failures have been eliminated from ABB fuel by introducing the Guardian TM grid. Intermediate flow mixers that improve the thermal hydraulic performance and the dimensional stability of the fuel has also been developed within ABB. (author)

ASN takes position in the in-service follow-up programs of primary and secondary loops of EdF's nuclear power plants

International Nuclear Information System (INIS)

2002-01-01

This decision from the French authority of nuclear safety (ASN) aims at fixing the conditions to be respected by Electricite de France (EdF) during its in-service follow-up programs for the monitoring and preventive maintenance of the primary and secondary cooling loops of EdF's PWR reactors. The components and the particular points to be controlled are listed in appendixes. (J.S.)

Nuclear criticality safety analysis for the traveller PWR fuel shipping package

Energy Technology Data Exchange (ETDEWEB)

Vescovi, P.J.; Kent, N.A.; Casado, C.A. [Westinghouse Electric Co., LLC, Columbia, SC (United States)]|[ENUSA Industrias Avanzadas SA, Madrid (Spain)

2004-07-01

The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. Two immediately noticeable features of the Traveller are that it carries a single fuel assembly instead of two as do other package designs, and that it has built-in moderator, which forms part of the flux-trap system. The criticality safety case shows that the Traveller satisfies both U.S. and IAEA licensing requirements, and demonstrates that the package remains acceptably subcritical under normal conditions and hypothetical accident conditions of transport. This paper looks at the modeling techniques that were used to analyze the several accident scenarios that were considered, including: Lattice pitch expansion; Lattice pitch expansion along the fuel assembly length; Preferential flooding (selective flooding of different cavities); Differential flooding (varying water levels inside different cavities); Partial flooding (varying water density); Axial rod displacement; o Sensitivity studies of variable foam densities and boron content in packaging; Analysis for carrying loose rods in a rodbox; The criticality safety case for the Traveller proved to be a successful cooperative effort between ENUSA and Westinghouse.

Nuclear criticality safety analysis for the traveller PWR fuel shipping package

International Nuclear Information System (INIS)

Vescovi, P.J.; Kent, N.A.; Casado, C.A.

2004-01-01

The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. Two immediately noticeable features of the Traveller are that it carries a single fuel assembly instead of two as do other package designs, and that it has built-in moderator, which forms part of the flux-trap system. The criticality safety case shows that the Traveller satisfies both U.S. and IAEA licensing requirements, and demonstrates that the package remains acceptably subcritical under normal conditions and hypothetical accident conditions of transport. This paper looks at the modeling techniques that were used to analyze the several accident scenarios that were considered, including: Lattice pitch expansion; Lattice pitch expansion along the fuel assembly length; Preferential flooding (selective flooding of different cavities); Differential flooding (varying water levels inside different cavities); Partial flooding (varying water density); Axial rod displacement; o Sensitivity studies of variable foam densities and boron content in packaging; Analysis for carrying loose rods in a rodbox; The criticality safety case for the Traveller proved to be a successful cooperative effort between ENUSA and Westinghouse

Radiation dose rates from commercial PWR and BWR spent fuel elements

International Nuclear Information System (INIS)

Willingham, C.E.

1981-10-01

Data on measurements of gamma dose rates from commercial reactor spent fuel were collected, and documented calculated gamma dose rates were reviewed. As part of this study, the gamma dose rate from spent fuel was estimated, using computational techniques similar to previous investigations into this problem. Comparison of the measured and calculated dose rates provided a recommended dose rate in air versus distance curve for PWR spent fuel

PREP-PWR-1.0: a WIMS-D/4 pre-processor code for the generation of data for PWR fuel assemblies

International Nuclear Information System (INIS)

Ball, G.

1991-06-01

The PREP-PWR-1.0 computer code is a substantially modified version of the PREWIM code which formed part of the original MARIA System (Report J.E.N. 543). PREP-PWR-1.0 is a comprehensive pre-processor code which generates input data for the WIMS-D/4.1 code (Report PEL 294) for PWR fuel assemblies, with or without control and burnable poison rods. This data is generated at various base and off-base conditions. The overall cross section generation methodology is described, followed by a brief overview of the model. Aspects of the base/off-base calculational scheme are outlined. Additional features of the code are described while the input data format of PREP-PWR-1.0 is listed. The sample problems and suggestions for further improvements to the code are also described. 2 figs., 2 tabs., 12 refs

The physical and chemical degradation of PWR fuel rods in severe accident conditions

International Nuclear Information System (INIS)

Parsons, P.D.; Mowat, J.A.S.; Dewhurst, D.W.F.; Hughes, T.E.

1983-01-01

An experimental study of the interaction between Zircaloy-4 cladding and UO 2 in PWR fuel rods heated to high temperatures with a negligible differential pressure across the cladding wall is described. The fuel rods were of dimensions appropriate to the 17x17 PWR fuel sub-assembly and were heated in a non-oxidising environment (vacuum) up to approx. 1850 deg. C either isothermally or through heating ramps. Observations were made concerning the extent and nature of the reaction zone between Zircaloy-4 and UO 2 over the temperature range 1500-1850 deg. C for times ranging from 1 min to 125 min. The location, morphology and the chemical composition of the phases formed are described along with the kinetics of their formation. (author)

Characteristics of several equilibrium fuel cycles of PWR

International Nuclear Information System (INIS)

Waris, Abdul; Sekimoto, Hiroshi

2001-01-01

This paper evaluated the influence of neutron spectrum on characteristics of several equilibrium fuel cycles of pressurized water reactor (PWR). In this study, five kinds of fuel cycles were investigated. Required uranium enrichment, required natural uranium amount, and toxicity of heavy metals (HMs) in spent fuel were presented for comparison. The results showed that the enrichment and the required amount of natural uranium decrease significantly with increasing number of confined heavy nuclides when uranium is discharged from the reactor. On the other hand, when uranium is totally confined, the enrichment becomes extremely high. The confinement of plutonium and minor actinides (MA) seems effective in reducing radio-toxicity of discharged wastes. By confining all heavy nuclides except uranium those three characteristics could be reduced considerably. For this fuel cycle the toxicity of HMs in spent fuel become nearly equal to or less than that of loaded uranium. (author)

Primary circuit leak detection an application on PWR vessel head penetrations

International Nuclear Information System (INIS)

Loisy, F.; Germain, J.L.; Chauvel, L.

1996-01-01

In 1991, cracks were discovered and localized in the lower part of certain vessel head adapters in EDF PWR units. While awaiting the replacement of the vessel heads in question, EDF developed systems to enable continuous monitoring of vessel head penetration, by means of early detection of leaks. One of these systems in based on detection of water vapour in a confined space above the vessel head. The efficiency of the measurement chain is particularly dependent on dilution of the leakage in the confined space prior TO entry in the sampling circuit. The detection threshold for this method is on the order of 1.2 liters/hour for a dilution rate of 1500 rate of 1500 m 3 /h and a dew point of 22 deg C. This system has now been in operation on three 1300-MW PWR units for three years, and has proved to function satisfactorily. (authors)

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)

Shielding design for PWR in France

International Nuclear Information System (INIS)

Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

1983-05-01

Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983

Reactivity and neutron emission measurements of highly burnt PWR fuel rod samples

International Nuclear Information System (INIS)

Murphy, M.F.; Jatuff, F.; Grimm, P.; Seiler, R.; Brogli, R.; Meier, G.; Berger, H.-D.; Chawla, R.

2006-01-01

Fuel rods with burnup values beyond 50 GWd/t are characterised by relatively large amounts of fission products and a high abundance of major and minor actinides. Of particular interest is the change in the reactivity of the fuel as a function of burnup and the capability of modern codes to predict this change. In addition, the neutron emission from burnt fuel has important implications for the design of transport and storage facilities. Measurements have been made of the reactivity effects and the neutron emission rates of highly burnt uranium oxide and mixed oxide fuel rod samples coming from a pressurised water reactor (PWR). The reactivity measurements have been made in a PWR lattice in the PROTEUS zero-energy reactor moderated in turn with: water, a water and heavy water mixture and water containing boron. A combined transport flask and sample changer was used to insert the 400 mm long burnt fuel rod segments into the reactor. Both control rod compensation and reactor period methods were used to determine the reactivities of the samples. For the range of burnup values investigated, an interesting exponential relationship has been found between the neutron emission rate and the measured reactivity

Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP

International Nuclear Information System (INIS)

Vargas E, S.; Esquivel E, J.; Ramirez S, J. R.

2013-10-01

The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

PWR and BWR spent fuel assembly gamma spectra measurements

Energy Technology Data Exchange (ETDEWEB)

Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

2016-10-11

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

Conceptual design study of small long-life PWR based on thorium cycle fuel

International Nuclear Information System (INIS)

Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul; Permana, Sidik

2014-01-01

A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of 233 U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWR result small excess reactivity and reduced power peaking during its operation

Decruding of PWR Fuel with the ICEDEC{sup TM} technique

Energy Technology Data Exchange (ETDEWEB)

Ivars, Roland; Fredriksson, Eva; Rosengren, Anders; Hallgren, Peter [Westinghouse Atom AB, Uppsala (Sweden)

2002-04-15

The novel fuel decontamination system ICEDEC{sup TM} utilizes a mixture of ice particles and water circulating through the assembly for scraping off loose crud from the fuel surfaces. Initially, ICEDEC{sup TM} was developed for BWRs. By means of fuel decontamination the origin of the personnel radiation exposure, i. e. the loose fuel crud, can be decreased. The technique can be used for radiation level reduction, eliminating the AOA (Axial Offset anomaly) problem or general decruding purposes. Decontamination tests of burned out and two-year-old fuel in the BWR Ringhals 1 in Sweden verified that ICEDEC{sup TM} maintained the mechanical integrity of the fuel and fulfilled the very important criteria; only the loose crud was removed and the oxide layer was not affected. Activity measurements prior to and after the decontamination, showed that more than 50% of the loose crud was removed from the fuel surfaces. For PWR the condition of maintained boron level in the pool water is a prerequisite. This can be achieved by utilizing a closed loop system with a water reservoir. From the reservoir water is used for the production of the ice particles. After removal of the loose crud from the fuel the crud/ice/water slurry is separated in a specially designed filter unit. The melted ice and residual water is then transported back to the water reservoir. Other methods could be to add boron equivalent to the excess water from the melted ice or use reverse osmosis to separate the boron and water in the pool. Including an application study followed by preliminary and detailed designs and manufacture and testing a PWR ICEDEC{sup TM} system can be licensed after two years.

Performance Specification Shippinpark Pressurized Water Reactor Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shippingport Spent Fuel Canisters

International Nuclear Information System (INIS)

JOHNSON, D.M.

2000-01-01

This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders

Siemens Nuclear Power Corporation experience with BWR and PWR fuels

International Nuclear Information System (INIS)

Reparaz, A.; Smith, M.H.; Stephens, L.G.

1992-01-01

The large data base of fuel performance parameters available to Siemens Nuclear Power Corporation (SNP), and the excellent track record of innovation and fuel reliability accumulated over the last twenty-three years, allows SNP to have a clear insight on the characteristics of future developments in the area of fuel design. Following is a description of some of SNP's recent design innovations to prevent failures and to extend burnup capabilities. A goal paramount to the design and manufacture of BWR and PWR fuel is that of zero defects from any case during its operation in the reactor. Progress has already been made in achieving this goal. This paper summarized the cumulative failure rate of SNP fuel rod through January 1992

Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

Energy Technology Data Exchange (ETDEWEB)

Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)

2002-08-01

The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

International Nuclear Information System (INIS)

Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir

2002-01-01

The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them 241 Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides 242 Cm and 244 Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile 239 Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of 241 Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% 238 Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

Chemical and radiochemical specifications - PWR power plants

International Nuclear Information System (INIS)

Stutzmann, A.

1997-01-01

Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

Multi-recycling of transuranic elements in a PWR assembly with reduced fuel rod diameter

International Nuclear Information System (INIS)

Chambers, Alex; Ragusa, Jean C.

2014-01-01

Highlights: • Study of multiple recycling passes of transuranic elements: (a) without exceeding 5 wt.% on U-235 enrichment; (b) using PWR fuel assemblies compatible with current reactor core internals. • Isotopic concentrations tend towards an equilibrium after 15 recycle passes, suggesting that thermal recycling may be continued beyond that point. • Radiotoxicity comparisons for once-through UOX, once-recycle MOX-Pu, and multiple recycle passes of MOX-PuNpAm and MOX-PuNpAmCm are presented. - Abstract: This paper examines the multi-recycling of transuranic (TRU) elements (Pu-Np-Am-Cm) in standard Pressurized Water Reactor (PWR) assemblies. The original feed of TRU comes from legacy spent UOX fuel. For all subsequent recycling passes, TRU elements from the previous generation are employed, supplemented by TRU from legacy UOX fuel, as needed. The design criteria include: 235 U enrichment requirements to remain below 5 w/o, TRU loading limits to avoid return to criticality under voided conditions, and assembly power peaking factors. In order to carry out multiple recycling passes within the design envelope, additional neutron moderation is required and achieved by reducing the fuel pellet diameter by about 13%, thus keeping the assembly design compatible with current PWR core internals. TRU transmutation rates and long-term ingestion radiotoxicity results are presented for 15 recycling passes and compared to standard UOX and MOX once-through cycles. The results also show that TRU fuel isotopics and radiotoxicity tend towards an equilibrium, enabling further additional recycling passes

Low-density moderation in the storage of PWR fuel assemblies

International Nuclear Information System (INIS)

Alcorn, F.M.

1987-01-01

The nuclear criticality safety of PWR fuel storage arrays requires that the potential of low-density moderation within the array be considered. The calculated criticality effect of low-density moderation in a typical PWR fuel assembly array is described in this paper. Calculated reactivity due to low-density moderation can vary significantly between physics codes that have been validated for well moderated systems. The availability of appropriate benchmark experiments for low-density moderation is quite limited; attempts to validate against the one set of suitable experiments at low density have been disappointing. Calculations indicate that a typical array may be unacceptable should the array be subjected to interstitial moderation equivalent to 5 % of full density water. Array parameters (such as spacing and size) will dramatically affect the calculated maximum K-eff at low-density moderation. Administrative and engineered control may be necessary to assure maintenance of safety at low-density moderation. Potential sources for low-density moderation are discussed; in general, accidentally achieving degrees of low-density moderation which might lead to a compromise of safety are not credible. (author)

Estimation of PWR spent fuel composition using SCALE and SWAT code systems

Energy Technology Data Exchange (ETDEWEB)

Shin, Hee Sung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kenya, Suyama; Hiroshi, Okuno [Japan Atomic Energy Research Institute, Tokyo (Japan)

2001-05-01

The isotopic composition calculations were performed for 26 spent fuel samples from Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using SCALE4.4 SAS2H with 27, 44 and 238 group cross-section libraries and SWAT with 107 group cross-section library. For convenience, the ratio of the measured to calculated value was used as a parameter. The four kinds of the calculation results were compared with the measured data. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed the following results. Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from Obrigheim reactor. Larger than unity ratios were found for Am-241 for both the 16 and 55 samples. Larger than unity ratios were found for Sm-149 for the 55 samples. In the case of 26 sample SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor of a system containing PWR spent fuel, taking burnup credit into account.

VANTAGE 5 PWR fuel assembly demonstration program at Virgil C. Summer nuclear station

International Nuclear Information System (INIS)

Warner, D.C.; Orr, W.L.

1985-01-01

VANTAGE 5 is an improved PWR fuel product designed and manufactured by Westinghouse Electric Corporation. The VANTAGE 5 fuel design features integral fuel burnable absorbers, intermediate flow mixer grids, axial blankets, high burnup capability, and a reconstitutable top nozzle. A demonstration program for this fuel design commenced in late 1984 in cycle 2 of the Virgil C. Summer Nuclear Station. Objectives for VANTAGE 5 fuel are reduced fuel cycle costs, better core operating margins, and increased design and operating flexibility. Inspections of the VANTAGE 5 demonstration assemblies are planned at each refueling outage

Preparation for Future Defuelling and Decommissioning Works on EDF Energy's UK Fleet of Advanced Gas Cooled Reactors

International Nuclear Information System (INIS)

Bryers, John; Ashmead, Simon

2016-01-01

EDF Energy/Nuclear Generation is the owner and operator of 14 Advanced Gas cooled Reactors (AGR) and one Pressurised Water Reactor (PWR), on 8 nuclear stations in the UK. EDF Energy/Nuclear Generation is responsible for all the activities associated with the end of life of its nuclear installations: de-fuelling, decommissioning and waste management. As the first AGR is forecast to cease generation within 10 years, EDF Energy has started planning for the decommissioning. This paper covers: - broad outline of the technical strategy and arrangements for future de-fuelling and decommissioning works on the UK AGR fleet, - high level strategic drivers and alignment with wider UK nuclear policy, - overall programme of preparation and initial works, - technical approaches to be adopted during decommissioning. (authors)

Methods and computer programs for PWR's fuel management: Programs Sothis and Ciclon

International Nuclear Information System (INIS)

Aragones, J.M.; Corella, M.R.; Martinez-Val, J.M.

1976-01-01

Methos and computer programs developed at JEN for fuel management in PWR are discussed, including scope of model, procedures for sistematic selection of alternatives to be evaluated, basis of model for neutronic calculation, methods for fuel costs calculation, procedures for equilibrium and trans[tion cycles calculation with Soth[s and Ciclon codes and validation of methods by comparison of results with others of reference (author) ' [es

Survey of the power ramp performance testing of KWU'S PWR UO 2, fuel

Science.gov (United States)

Ga¨rtner, M.; Fischer, G.

1987-06-01

To determine the power ramp performance of KWU's PWR UO 2 fuel, 134 fuel rodlets with burnups of up to 46 GWd/ t (U) and several fuel assemblies with 19 to 30 GWd/t (U) burnup were ramped in power in the research reactors HFR Petten/The Netherlands and R2 Studsvik/Sweden and in the power plants KWO and KWB-A/Germany, respectively. The power ramp tests demonstrate decreasing resistance of the PWR fuel rods to PCI (pellet-to-clad interaction) up to fuel burnups of 35 GWd/t (U) and a reversal effect at higher burnups. The fuel rods can be operated free of defects at fast power transients to linear heat generation rates of up to 400 W/cm, at least.Power levels of up to 490 W/cm can be reached without defects by reducing the ramp rate. After reshuffling according to an out-in scheme, 1-cycle fuel assemblies may return to rod powers of up to 480 W/cm with a power increase rate of up to 10 W/(cm min) without fuel rod damage. Set points basing on these test results and incorporated into the power distribution control and power density limitation system of KWU's advanced power plants guarantee safe plant operation under normal and load follow operating conditions.

Development of the vacuum drying process for the PWR spent nuclear fuel dry storage

Energy Technology Data Exchange (ETDEWEB)

Baeg, Chagn Yeal; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

2016-12-15

This paper describes the development of a dry operation process for PWR spent nuclear fuel, which is currently stored in the domestic NPP's storage pool, using a dual purpose metal cask. Domestic NNPs have had experience with wet type transportation of PWR spent nuclear fuel between neighboring NPPs since the early 1990s, but no experience with dry type operation. For this reason, we developed a specific operation process and also confirmed the safety of the major cask components and its spent nuclear fuel during the dual purpose metal cask operation process. We also describe the short term operation process that was established to be completed within 21 hours and propose the allowable working time for each step (15 hours for wet process, 3 hours for drain process and 3 hours for vacuum drying process)

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

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

International Nuclear Information System (INIS)

Boussard, F.; Huillery, R.

1994-01-01

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

Fuel Cycle Cost Calculations for a 120,000 shp PWR for Ship Propulsion. RCN Report

International Nuclear Information System (INIS)

Dekker, N.H.; Foggi, C.; Giacomazzi, G.

1972-02-01

A parametric study of the fuel cycle costs for a 120,000 SHP PWR for ship propulsion has been carried out. Variable parameters are: fuel pellet diameter, moderating ratio and refuelling scheme. Minimum fuel cycle costs can be obtained at moderating ratios of about 2.2. Both fuel cycle costs and reactor control requirements favour the two batch core. (author)

Multi-recycling of transuranic elements in a PWR assembly with reduced fuel rod diameter

Energy Technology Data Exchange (ETDEWEB)

Chambers, Alex, E-mail: acchamb@gmail.com; Ragusa, Jean C., E-mail: jean.ragusa@tamu.edu

2014-04-01

Highlights: • Study of multiple recycling passes of transuranic elements: (a) without exceeding 5 wt.% on U-235 enrichment; (b) using PWR fuel assemblies compatible with current reactor core internals. • Isotopic concentrations tend towards an equilibrium after 15 recycle passes, suggesting that thermal recycling may be continued beyond that point. • Radiotoxicity comparisons for once-through UOX, once-recycle MOX-Pu, and multiple recycle passes of MOX-PuNpAm and MOX-PuNpAmCm are presented. - Abstract: This paper examines the multi-recycling of transuranic (TRU) elements (Pu-Np-Am-Cm) in standard Pressurized Water Reactor (PWR) assemblies. The original feed of TRU comes from legacy spent UOX fuel. For all subsequent recycling passes, TRU elements from the previous generation are employed, supplemented by TRU from legacy UOX fuel, as needed. The design criteria include: {sup 235}U enrichment requirements to remain below 5 w/o, TRU loading limits to avoid return to criticality under voided conditions, and assembly power peaking factors. In order to carry out multiple recycling passes within the design envelope, additional neutron moderation is required and achieved by reducing the fuel pellet diameter by about 13%, thus keeping the assembly design compatible with current PWR core internals. TRU transmutation rates and long-term ingestion radiotoxicity results are presented for 15 recycling passes and compared to standard UOX and MOX once-through cycles. The results also show that TRU fuel isotopics and radiotoxicity tend towards an equilibrium, enabling further additional recycling passes.

Physics of plutonium and americium recycling in PWR using advanced fuel concepts

International Nuclear Information System (INIS)

Hourcade, E.

2004-01-01

PWR waste inventory management is considered in many countries including Frances as one of the main current issues. Pu and Am are the 2 main contents both in term of volume and long term radio-toxicity. Waiting for the Generation IV systems implementation (2035-2050), one of the mid-term solutions for their transmutation involves the use of advanced fuels in Pressurized Water Reactors (PWR). These have to require as little modification as possible of the core internals, the cooling system and fuel cycle facilities (fabrication and reprocessing). The first part of this paper deals with some neutronic characteristics of Pu and/or Am recycling. In a second part, 2 technical solutions MOX-HMR and APA-DUPLEX-84 are presented and the third part is devoted to the study of a few global strategies. The main neutronic parameters to be considered for Pu and Am recycling in PWR are void coefficient, Doppler coefficient, fraction of delayed neutrons and power distribution (especially for heterogeneous configurations). The modification of the moderation ratio, the opportunity to use inert matrices (targets), the optimisation of Uranium, Plutonium and Americium contents are the key parameters to play with. One of the solutions (APA-DUPLEX-84) presented here is a heterogeneous assembly with regular moderation ratio composed with both target fuel rods (Pu and Am embedded in an inert matrix) and standard UO 2 fuel rods. An EPR (European Pressurised Reactor) type reactor, loaded only with assemblies containing 84 peripheral targets, can reach an Americium consumption rate of (4.4; 23 kg/TWh) depending on the assembly concept. For Pu and Am inventories stabilisation, the theoretical fraction of reactors loaded with Pu + Am or Pu assemblies is about 60%. For Americium inventory stabilisation, the fraction decreases down to 16%, but Pu is produced at a rate of 18.5 Kg/TWh (-25% compared to one through UOX cycle)

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)

Zinc injection on the EDF pressurized light water reactors. Current results and operating experience feedback

International Nuclear Information System (INIS)

Piana, Olivier; Duval, Arnaud; Moleiro, Edgar; Benfarah, Moez; Bretelle, Jean-Luc; Chaigne, Guy

2014-01-01

Nowadays, zinc injection, as well as pH management and hydrogen control, is increasingly considered as an essential element of PWR Primary Water Chemistry worldwide. After a first implementation of zinc injection at Bugey 2 since 2004 and Bugey 4 since 2006, EDF decided to extend this practice, which constitutes a modification of primary circuit chemical conditioning, to other units of its fleet. Currently, 15 among the 58 reactors of the French fleet are injecting depleted zinc acetate into the primary coolant water. Three main goals were identified at the beginning of this program. Indeed, the expected benefits of zinc injection were: Reduction of the rate of generalized corrosion and mitigation of stress corrosion cracking initiation on nickel based alloys (Material goal). Curative or preventive reduction of radiation sources to which workers are exposed (Radiation fields' goal). Mitigation of the AOA or CIPS risks by reduction of corrosion products releases and mitigation of crud deposition (Fuel protection goal). To monitor the zinc addition, EDF has defined a complete survey program concerning: chemistry and radiochemistry responses (primary coolant monitoring of corrosion and fission products and calculation of zinc injected, zinc removed and zinc incorporated in RCS surfaces) ; radiation fields (dose rates and deposited activities measurements) ; materials (statistical analysis of SG tube cracks) ; fuel (oxide thickness measurements and visual exams) ; effluents (corrosion products releases and isotopic distribution follow up) ; wastes (radiochemical characterization of filters). This paper will detail the present results of this monitoring program. It appears that the expected benefits of zinc injection have yet to be fully realized; further operating experience will be required in order to fully evaluate its impact. (author)

Modeling of PWR fuel at extended burnup

Energy Technology Data Exchange (ETDEWEB)

Dias, Raphael M.; Silva, Antonio Teixeira, E-mail: rmdias@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

Modeling of PWR fuel at extended burnup

International Nuclear Information System (INIS)

Dias, Raphael M.; Silva, Antonio Teixeira

2015-01-01

Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

EDF experience with open-quotes hot spotclose quotes management

International Nuclear Information System (INIS)

Guio, J.M. de

1995-01-01

During the past few years, open-quotes hot spotsclose quotes due to the presence of particles of metal activated during their migration through the reactor core, have been detected at several French pressurized water reactor (PWR) units. These open-quotes hot spots,close quotes which generate very high dose rates (from about 10 Gy/h to 200 G/h) are a significant factor in increase occupational exposures during outrates. Of particular concern are the difficult cases which prolong outage duration and increase the volume of radiological waste. Confronted with this situation, Electricite de France (EDF) has set up a national research group, as part of its ALARA program, to establish procedures and techniques to avoid, detect, and eliminate of hot spots. In particular, specific processes have been developed to eliminate these hot spots which are most costly in terms of occupational exposure due to the need for reactor maintenance. This paper sets out the general approach adopted at EDF so far to cope with the problem of hot spots, illustrated by experience at Blayais 3 and 4

PWR fuel monitoring: recent progress with hot cells' examination equipment

International Nuclear Information System (INIS)

Chenebault, P.

1989-01-01

The 'hot' laboratories set up by the French Atomic Energy Authority (CEA) in its nuclear research centers at Saclay and Grenoble, and by the French Electricity Board (EDF) on the Chinon nuclear power station site, are used for dismantling and examining fuel rod assemblies irradiated in PWRs. This article is limited to a description of a number of new or totally updated items of equipment in these laboratories. Nuclear industry companies are also participating in the development of new examination techniques. As an example, the use of wave-guides for remote transmission of signals in a radioactive environment is described. 2 figs

Optimization of PWR fuel assembly radial enrichment and burnable poison location based on adaptive simulated annealing

International Nuclear Information System (INIS)

Rogers, Timothy; Ragusa, Jean; Schultz, Stephen; St Clair, Robert

2009-01-01

The focus of this paper is to present a concurrent optimization scheme for the radial pin enrichment and burnable poison location in PWR fuel assemblies. The methodology is based on the Adaptive Simulated Annealing (ASA) technique, coupled with a neutron lattice physics code to update the cost function values. In this work, the variations in the pin U-235 enrichment are variables to be optimized radially, i.e., pin by pin. We consider the optimization of two categories of fuel assemblies, with and without Gadolinium burnable poison pins. When burnable poisons are present, both the radial distribution of enrichment and the poison locations are variables in the optimization process. Results for 15 x 15 PWR fuel assembly designs are provided.

RCC-C: Design and construction rules for fuel assemblies of PWR nuclear power plants

International Nuclear Information System (INIS)

2015-01-01

The RCC-C code contains all the requirements for the design, fabrication and inspection of nuclear fuel assemblies and the different types of core components (rod cluster control assemblies, burnable poison rod assemblies, primary and secondary source assemblies and thimble plug assemblies). The design, fabrication and inspection rules defined in RCC-C leverage the results of the research and development work pioneered in France, Europe and worldwide, and which have been successfully used by industry to design and build nuclear fuel assemblies and incorporate the resulting feedback. The code's scope covers: fuel system design, especially for assemblies, the fuel rod and associated core components, the characteristics to be checked for products and parts, fabrication methods and associated inspection methods. The RCC-C code is used by the operator of the PWR nuclear power plants in France as a reference when sourcing fuel from the world's top two suppliers in the PWR market, given that the French operator is the world's largest buyer of PWR fuel. Fuel for EPR projects is manufactured according to the provisions of the RCC-C code. The code is available in French and English. The 2005 edition has been translated into Chinese. Contents of the 2015 edition of the RCC-C code: Chapter 1 - General provisions: 1.1 Purpose of the RCC-C, 1.2 Definitions, 1.3 Applicable standards, 1.4 Equipment subject to the RCC-C, 1.5 Management system, 1.6 Processing of non-conformances; Chapter 2 - Description of the equipment subject to the RCC-C: 2.1 Fuel assembly, 2.2 Core components; Chapter 3 - Design: Safety functions, operating functions and environment of fuel assemblies and core components, design and safety principles; Chapter 4 - Manufacturing: 4.1 Materials and part characteristics, 4.2 Assembly requirements, 4.3 Manufacturing and inspection processes, 4.4 Inspection methods, 4.5 Certification of NDT inspectors, 4.6 Characteristics to be inspected for the

Radionuclide release from PWR spent fuel specimens with induced cladding defects

International Nuclear Information System (INIS)

Wilson, C.N.; Oversby, V.M.

1984-03-01

Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented

Audit on modalities of assessment of gross claims for the calculation of provisions of deconstruction of currently operated EDF reactors. Report synthesis, August 4, 2015

International Nuclear Information System (INIS)

2015-01-01

Based on data provided by EDF, this report presents hypotheses, calculation methods and analyses regarding modalities of assessment of provisions for the deconstruction of EDF reactors. After a presentation of the EDF PWR fleet, an overview of dismantling activities at the world scale, a discussion of accounting provisions, a presentation of the audit methodology and limitations, the authors report an analysis of deconstruction costs through the case of Dampierre (dismantling strategy, planning, engineering expenses, site-related expenses, dismantling costs, demolition costs, waste management cost). They propose a comparison at the international level, discuss an extrapolation to the whole fleet, evoke a risk analysis, and presents their conclusion for the global assessment

Control rod ejection accident analysis for a PWR with thorium fuel loading

Energy Technology Data Exchange (ETDEWEB)

Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)

2010-07-01

This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

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)

Fuel assembly for pressure loss variable PWR type reactor

International Nuclear Information System (INIS)

Yoshikuni, Masaaki.

1993-01-01

In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

The thermo-mechanics of the PWR fuel rod; La thermomecanique du crayon de combustible REP

Energy Technology Data Exchange (ETDEWEB)

Barral, J.C. [Electricite de France, EDF, 75 - Paris (France); Gautier, B.; Chaigne, G. [Electricite de France, Service Etudes et Projets Thermiques et Nucleaires, 75 - Paris (France)] [and others

1999-03-29

The fuel rod mechanics is of a great importance in the safety and performance of the reactors. In this domain a meeting has been organized by the SFEN the 18 march 1998 at Paris. With the participation of scientists from CEA, EDF and Framatome, the physics of the fuel rods was presented based on four main aspects. Two first papers dealt with the solicitations of the fuel rod in normal and accidental conditions. The physical phenomena under irradiation were then detailed in the four following talks. Three papers presented the simulation and the codes of the fuel-cladding interactions with the diabolo effect. The last paper was devoted to the experiment feedback and the research programs. (A.L.B.)

Radionuclide release from PWR spent fuel specimens with induced cladding defects

International Nuclear Information System (INIS)

Wilson, C.N.; Oversby, V.M.

1984-03-01

Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented. 6 references, 4 figures

AREVA's fuel assemblies addressing high performance requirements of the worldwide PWR fleet

International Nuclear Information System (INIS)

Anniel, Marc; Bordy, Michel-Aristide

2009-01-01

Taking advantage of its presence in the fuel activities since the start of commercial nuclear worldwide operation, AREVA is continuing to support the customers with the priority on reliability, to: >participate in plant operational performance for the in core fuel reliability, the Zero Tolerance for Failure ZTF as a continuous improvement target and the minimisation of manufacturing/quality troubles, >guarantee the supply chain a proven product stability and continuous availability, >support performance improvements with proven design and technology for fuel management updating and cycle cost optimization, >support licensing assessments for fuel assembly and reloads, data/methodologies/services, >meet regulatory challenges regarding new phenomena, addressing emergent performance issues and emerging industry challenges for changing operating regimes. This capacity is based on supplies by AREVA accumulating very large experience both in manufacturing and in plant operation, which is demonstrated by: >manufacturing location in 4 countries including 9 fuel factories in USA, Germany, Belgium and France. Up to now about 120,000 fuel assemblies and 8,000 RCCA have been released to PWR nuclear countries, from AREVA European factories, >irradiation performed or in progress in about half of PWR world wide nuclear plants. Our optimum performances cover rod burn ups of to 82GWD/tU and fuel assemblies successfully operated under various world wide fuel management types. AREVA's experience, which is the largest in the world, has the extensive support of the well known fuel components such as the M5'TM'cladding, the MONOBLOC'TM'guide tube, the HTP'TM' and HMP'TM' structure components and the comprehensive services brought in engineering, irradiation and post irradiation fields. All of AREVA's fuel knowledge is devoted to extend the definition of fuel reliability to cover the whole scope of fuel vendor support. Our Top Reliability and Quality provide customers with continuous

PWR: 10 years after and perspectives

International Nuclear Information System (INIS)

1990-01-01

These proceedings of the SFEN days on PWR (Ten years after and perspectives) comprise 13 conferences bearing on: - From the occurential approach to the state approach - Evolution of calculating tools - Human factors and safety - Reactor safety in the PWR 2000 - The PWR and the electrical power grid load follow - Fuel aspect of PWR management - PWR chemistry evolution - Balance of radiation protection - PWR modifications balance and influence on reactor operation - Design and maintenance of reactor components: 4 conferences [fr

Computer code TOBUNRAD for PWR fuel bundle heat-up calculations

International Nuclear Information System (INIS)

Shimooke, Takanori; Yoshida, Kazuo

1979-05-01

The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)

EDF experience with {open_quotes}hot spot{close_quotes} management

Energy Technology Data Exchange (ETDEWEB)

Guio, J.M. de [Blayais Nuclear Power Plant, St. Ciers (France)

1995-03-01

During the past few years, {open_quotes}hot spots{close_quotes} due to the presence of particles of metal activated during their migration through the reactor core, have been detected at several French pressurized water reactor (PWR) units. These {open_quotes}hot spots,{close_quotes} which generate very high dose rates (from about 10 Gy/h to 200 G/h) are a significant factor in increase occupational exposures during outrates. Of particular concern are the difficult cases which prolong outage duration and increase the volume of radiological waste. Confronted with this situation, Electricite de France (EDF) has set up a national research group, as part of its ALARA program, to establish procedures and techniques to avoid, detect, and eliminate of hot spots. In particular, specific processes have been developed to eliminate these hot spots which are most costly in terms of occupational exposure due to the need for reactor maintenance. This paper sets out the general approach adopted at EDF so far to cope with the problem of hot spots, illustrated by experience at Blayais 3 and 4.

Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

Energy Technology Data Exchange (ETDEWEB)

Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2013-10-15

The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

Construction and utilization of linear empirical core models for PWR in-core fuel management

International Nuclear Information System (INIS)

Okafor, K.C.

1988-01-01

An empirical core-model construction procedure for pressurized water reactor (PWR) in-core fuel management is developed that allows determining the optimal BOC k ∞ profiles in PWRs as a single linear-programming problem and thus facilitates the overall optimization process for in-core fuel management due to algorithmic simplification and reduction in computation time. The optimal profile is defined as one that maximizes cycle burnup. The model construction scheme treats the fuel-assembly power fractions, burnup, and leakage as state variables and BOC zone enrichments as control variables. The core model consists of linear correlations between the state and control variables that describe fuel-assembly behavior in time and space. These correlations are obtained through time-dependent two-dimensional core simulations. The core model incorporates the effects of composition changes in all the enrichment control zones on a given fuel assembly and is valid at all times during the cycle for a given range of control variables. No assumption is made on the geometry of the control zones. A scatter-composition distribution, as well as annular, can be considered for model construction. The application of the methodology to a typical PWR core indicates good agreement between the model and exact simulation results

General review of diagnostic systems in EDF PWR units

International Nuclear Information System (INIS)

Chevalier, R.; Brasseur, S.; Ricard, B.

1998-01-01

Since the beginning of the French nuclear program, Electricite de France (EDF) has looked for ways to improve the availability and safety of its nuclear units. Therefore, monitoring systems on turbogenerators, reactor coolant pumps, primary circuits and core internal structures were designed by the Research and Development Division and implemented with technologies available during the 1970's. However, mainly because of difficulties for data interpretation by plant personnel, EDF subsequently decided to design and develop different tools to help plant operators to process a diagnosis: - a new generation of the Monitoring and Diagnostic System called PSAD, - expert systems for diagnosis on reactor coolant pumps (RCP) 'DIAPO' and turbogenerator units (TG) 'DIVA', - diagnostic guides written for most equipment pending the installation of new monitoring and diagnosis systems. The first version of PSAD, installed in five units, performs on-line monitoring of the turbogenerator shaft line, steam inlet valves, the reactor coolant pumps and the generator stator. The second version not yet implemented, will include Loose Part Detection (LPD) and Core Internal Structure Monitoring (CISM). The level of diagnosis achieved by PSAD depends on the component monitored. The TG and RCP monitoring functions of PSAD compute high level diagnosis descriptors such as natural frequencies and long term trends but do not elaborate a diagnosis automatically. However, a diagnostic assistance window is available on-line, whenever a warning message is displayed, whether for immediate or later action. The window presents a diagnostic approach whose purpose is to find the causes of the symptoms observed. This diagnosis approach is automated in the DIVA and DIAPO expert systems. Numerous potential faults can be identified for both systems thanks to a hierarchy of abnormal situations. The user interactively answers questions when information is needed to progress in the diagnosis. The resulting

Secure and effective valve stem sealing in PWR power generating plants

International Nuclear Information System (INIS)

Reynolds, J.

1991-01-01

The PWR power generating plant combines severe operating conditions with the highest safety requirements, making it one of the most demanding environments for seals. An analysis of the conditions inherent in its operation reveals: an aggressive and radioactive fluid at high temperature and pressure; frequent thermal shocks; and hazards for maintenance personnel in the containment area unless the reactor is shut down. The achievement of today's quality and safety standards owes much to the experience, research and testing carried out by the Electricite de France during its graduation from its first nuclear unit to become the world's most important manager of PWR plants with over 45 now under its control. The number of valves involved in the French nuclear program is in excess of 1,300,000. Knowing what the affect of a leak can be, especially if it necessitates a shutdown of the power station, the need to insure the quality of valve sealing can be appreciated. At the beginning of their nuclear building program, the EdF was finding that valves, representing only 2 percent of the investment in a PWR plant, caused 20% of the unwanted outages and cost 60% of the total of plant maintenance. In this report, the author endeavors to show how this problem was solved by team work and concerted action by the EdF, the valve constructors and seal manufacturer, not forgetting the importance of informing and training the maintenance and repair teams within the power stations themselves

Analysis of bubble pressure in the rim region of high burnup PWR fuel

Energy Technology Data Exchange (ETDEWEB)

Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-02-01

Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

Nupec thermal hydraulic test to evaluate post-DNB characteristics for PWR fuel assemblies (1. general test plan and results)

International Nuclear Information System (INIS)

Norio, Kono; Kenji, Murai; Kaichiro, Misima; Takayuki, Suemura; Yoshiei, Akiyama; Keiichi, Hori

2001-01-01

In the present thermal hydraulic design of Pressurized Water Reactor (PWR), a departure from nucleate boiling (DNB) under anticipated transient conditions is not allowed. However, it is recognized that the DNB dose not cause a fuel rod failure immediately, and a suitable reactor trip can prevent the core from severe damages. If the fuel rod temperature under the post-DNB conditions can be accurately evaluated, the potentially existing margin in the present design method will be quantitatively assessed. To establish the heat transfer evaluation method on post-DNB event for PWR thermal hydraulic design, Nuclear Power Engineering Corporation (NUPEC) started a program, NUPEC Thermal Hydraulic Test to Evaluate Post-DNB Characteristics for PWR Fuel Assemblies (NUPEC-TH-P), in 1995 (hereinafter the year means fiscal year) under the sponsorship of Ministry of Economy, Trade and industry (METI). This program is now under going until 2001. This paper is to show the overall plan and the status of NUPEC-TH-P. (authors)

Study on new-type fuel-related assembly handling tools for PWR NPP

International Nuclear Information System (INIS)

Fan Xiumei

2013-01-01

This article describes the design and study on a set of new-type fuel-related assembly snatching tools used for PWR NPP. The purpose is mainly to enhance the tool safety, reliability and convenientness by improvement of the mechanism and structure of the tool for snatching preciseness and avoiding from falling and abrasion of fuel-related assemblies for any condition. The new-type fuel-related assembly handling tools are compared with similar equipment in worldwide in terms of function, main technical characteristic, and safety and protection, some of them are better than the similar equipment in that they have reliable loading and unloading and conveying capabilities. (author)

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

International Nuclear Information System (INIS)

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

1981-11-01

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

SIROCCO project: 15 advanced instructor desk and 4 simulated control room for 900MW and 1300MW EDF power plant simulators

International Nuclear Information System (INIS)

Alphonse, J.; Roth, P.; Sicard, Y.; Rudelli, P.

2006-01-01

This presentation describes the fifteen advanced instructors station and four simulated control delivered to EDF in the frame of the SIROCCO project by the Consortium formed by ATOS Origin, CORYS Tess, for the Electricite de France (EDF). These instructor stations are installed on fifteen replica training simulators located on different sites throughout France for the purposes of improving the job-related training of the EDF PWR nuclear power plant operating teams. This covers all 900 MW and 1300MW nuclear power plant of EDF. The simulated control rooms are installed on maintenance platform located at EDF and the consortium facilities. The consortium uses it to maintain and upgrade the simulators. EDF uses it to validate the upgrade delivered by the consortium before on site installation and to perform engineering analysis. This presentation sets out successively: - The major advantages of the generic and configurable connected module concept for flexible and quick adaptation to different simulators; - The innovative functionalities of the advanced Instructor Desk (IS) which make the instructor's tasks of preparation, monitoring and postanalysis of a training session easier and more homogeneous; - The use of the Simulated Control Room (SCR) for training purposes but also for those of maintenance and design studies for upgrades of existing control rooms

Evaluations on power ramp data of PWR fuels by FROST and THERMOST codes

International Nuclear Information System (INIS)

Murai, K.; Ogawa, S.; Nuno, H.; Kondo, Y.

1987-01-01

An evaluation is presented of power ramp data of Mitsubishi's PWR fuel rods tested in R-2, Studsvik, which was analysed by FROST and THERMOST codes. The analyses give good predictions for measured diameter changes and on-power rod elongations. The work indicates that FROST is capable of analysing both radial and axial pellet-cladding mechanism interaction (PCMI) appropriately, and that predicted states of PCMI (i.e. stress and strain which cannot be measured directly) are considered to be reliable. The ramp data used in the present analyses were obtained in two joint programmes with five Japanese PWR utilities (KEPCO, KYEPCO, SEPCO, HEPCO, and JAPCO). (UK)

Radioprotection and safety for a dry storage module for bare PWR fuel elements

International Nuclear Information System (INIS)

Tzontlimatzin, E.

1983-01-01

A module for dry storage of spent fuel from PWR, after a previous cooling time of 2 years, is examined. Biological protection is obtained by 185 cm of concrete. The safety study shows the impossibility of a fast increase in temperature in case of cooling system failure because in this case the module will be cooled by natural convection or thermosiphon. A project for a storage installation consisting of 5 modules for 1500 irradiated fuel assemblies is described [fr

PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements

International Nuclear Information System (INIS)

Ganzmann, I.; Hille, D.; Staude, U.

2009-01-01

The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

Corrosion behaviour of Zircaloy 4 fuel cans for high burnup in EdF PWRs

International Nuclear Information System (INIS)

Blat, M.; Kerrec, O.; Bourgoin, J.; Vrignaud, E.; Amanrich, H.

1994-01-01

Uniform corrosion of fuel cladding could be a limitation for burn-up enhancement. First, the oxide thickness measured on fuel cladding for high burn-up has been compared to the prediction of the EDF code, CYRANO 2E. A comparative metallurgical characterization has been also performed on samples which were oxidized in pile and in autoclave. Then, laboratories studies have been launched for a better understanding of the corrosion mechanisms. A reflection was proposed on the two main theoretical concepts proposed for these mechanisms. Their kinetics could be controlled by transfers in liquid medium (electrolyte) or in solid medium (compact oxide). For the first topic, a nanoscopic characterization of the oxide is in progress, using Atomic Force Microscope. The first results are presented. In the second case, an electrochemical approach (impedance spectroscopy and voltametry) is developed in our laboratories. The obtained results could give some new keys in order to understand the influence of some parameters (alloys composition, coolant chemistry,...). (authors). 7 figs., 1 tab., 7 refs

Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

International Nuclear Information System (INIS)

Faure, M.; Marchand, L.

1985-02-01

The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation [fr

PWR AXIAL BURNUP PROFILE ANALYSIS

International Nuclear Information System (INIS)

J.M. Acaglione

2003-01-01

The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

International Nuclear Information System (INIS)

Dias, A.F.V.

1980-10-01

Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.) [pt

Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes

International Nuclear Information System (INIS)

Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

2016-01-01

To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

EDF's experience with supercomputing and challenges ahead - towards multi-physics and multi-scale approaches

International Nuclear Information System (INIS)

Delbecq, J.M.; Banner, D.

2003-01-01

Nuclear power plants are a major asset of the EDF company. To remain so, in particular in a context of deregulation, competitiveness, safety and public acceptance are three conditions. These stakes apply both to existing plants and to future reactors. The purpose of the presentation is to explain how supercomputing can help EDF to satisfy these requirements. Three examples are described in detail: ensuring optimal use of nuclear fuel under wholly safe conditions, understanding and simulating the material deterioration mechanisms and moving forward with numerical simulation for the performance of EDF's activities. In conclusion, a broader vision of EDF long term R and D in the field of numerical simulation is given and especially of five challenges taken up by EDF together with its industrial and scientific partners. (author)

Reactivity and neutron emission measurements of burnt PWR fuel rod samples in LWR-PROTEUS phase II

International Nuclear Information System (INIS)

Murphy, M. F.; Jatuff, F.; Grimm, P.; Seiler, R.; Brogli, R.; Meier, G.; Berger, H. D.; Chawla, R.

2004-01-01

Measurements have been made of the reactivity effects and the neutron emission rates of uranium oxide and mixed oxide burnt fuel samples having a wide range of burnup values and coming from a Pressurised Water Reactor (PWR). The reactivity measurements have been made in a PWR lattice moderated in turn with: water, a water and heavy water mixture, and water containing boron. An interesting relationship has been found between the neutron emission rate and the measured reactivity. (authors)

Zinc injection on the EDF fleet monitoring the injection on 12 units

International Nuclear Information System (INIS)

Le-Meur, Gaelle Harmand; Anne-Marie; Stutzmann, Agnes; Taunier, Stephane; Benfarah, Moez; Bretelle, Jean-Luc; Alain, Rocher; Claeys, Myriam; Bonne, Sebastien

2012-09-01

After a first implementation of zinc injection at Bugey 2 and Bugey 4, EDF decided to extend the program to other units of its fleet. 14 more reactors from the French fleet of 58 were chosen in order to - Reduce the radiation sources for curative or preventive (after SGR) reasons - Mitigate stress corrosion cracking on nickel alloys and reduce the rate of generalized corrosion - Prevent the risk of CIPS, mainly after a fuel management change. Zinc injection started on 9 new units in 2011, 1 unit in 2012 and will be extended to 4 other units before the end of 2013. To monitor the injection, EDF has defined a complete program concerning chemistry, radiation protection (dose rate and deposited activities measurements), materials (statistical analysis of SG tube cracks), fuel (oxide measurements) and waste (radiochemical characterization of filters). Reference units were chosen for each field because of the size of the fleet. This paper will detail the different monitoring programs on the EDF plants injecting zinc. (authors)

EDF reactor building containment: Monitoring of the pre-stressed concrete structure

International Nuclear Information System (INIS)

Badez, N.

2009-01-01

The concrete containments of the EDF PWR are pre-stressed, and are monitored to observe the ageing effects on the structure, in particular the evolutions of creep, shrinkage, pre-stress loss, and air leakage tightness. Monitoring devices are installed during construction period, and measurements are checked, stored on a data base, and analysed during all the plant operating life time. The topic of the presentation is to present each part of the EDF monitoring organisation. A continuous monitoring makes it possible to produce periodical comprehensive reports about the mechanical analysis of the structure, the strain stabilisation,... Periodical tests (each 10 years) are planned. They consist to submit the containment to an internal air pressure at the accidental pressure level. The monitoring system gives the strain values in order to check their linearity and reversibility with decreasing pressure. At the same time, the containment tightness is checked with a specific instrumentation to verify that leak rate is lower than the required level. A general view of instrumentation implemented on the containment (sensors, data acquisition), and a data analysis are presented

Development of an advanced 16x165 Westinghouse type PWR fuel assembly for Slovenia

International Nuclear Information System (INIS)

Boone, M. L.; King, S. J.; Pulver, E. F.; Jeon, K.-L.; Esteves, R.; Kurincic, B.

2004-01-01

Industrias Nucleares do Brasil (INB), KEPCO Nuclear Fuel Company, Ltd. (KNFC), and Westinghouse Electric Company (Westinghouse) have jointly designed an advanced 16x16 Westinghouse type PWR fuel assembly. This advanced 16x16 Westinghouse type PWR fuel assembly, which will be implemented in both Kori Unit 2 (in Korea) and Angra Unit 1 (in Brazil) in January and March 2005, respectively, is an integral part of the utilities fuel management strategy. This same fuel design has also been developed for future use in Krsko Unit 1 (in Slovenia). In this paper we will describe the front-end nuclear fuel management activities utilized by the joint development team and describe how these activities played an integral part in defining the direction of the advanced 16x16 Westinghouse type PWR fuel assembly design. Additionally, this paper will describe how this design demonstrates improved margins under high duty plant operating conditions. The major reason for initiating this joint development program was to update the current 16x16 fuel assembly, which is also called 16STD. The current 16STD fuel assembly contains a non-optimized fuel rod diameter for the fuel rod pitch (i.e. 9.5 mm OD fuel rods at a 0.485 inch pitch), non-neutronic efficient components (i.e. Inconel Mid grids), no Intermediate Flow Mixer (IFM) grids, and other mechanical features. The advanced 16x16 fuel assembly is being designed for peak rod average burnups of up to 75 MWd/kgU and will use an optimized fuel rod diameter (i.e. 9.14 mm OD ZIRLO TM fuel rods), neutronic efficient components (i.e. ZIRLO TM Mid grids), ZIRLO TM Intermediate Flow Mixer (IFM) grids to improve Departure from Nucleate Boiling (DNB) margin, and many other mechanical features that improve design margins. Nuclear design activities in the areas of fuel cycle cost and fuel management were performed in parallel to the fuel assembly design efforts. As the change in reactivity due to the change in the fuel rod diameter influences directly

EDF, a utility and its own needs in the field of transport of nuclear materials

International Nuclear Information System (INIS)

Gouin, P.; Mignot, E.; Hoang, L.P.

1989-01-01

As one of the most important producers of nuclear electricity in the world, EDF is concerned by all the aspects of the transport of nuclear materials and more particularly by those related to the nuclear fuel cycle. EDF is not itself a specialist in this field and most of the transports along the nuclear fuel cycle is done for their own account by their usual partners such as COGEMA or TRANSNUCLEAIRE. Since the beginning of the French nuclear program, they have generally used for these transports casks that already exist on the market and which were well suited to their needs. Nevertheless, new and specific needs appeared during the progress of their nuclear program and have lead them to: study and build new casks or packages, use existing casks for new purposes, develop a device for the measurement of fuel assemblies burn up, develop a software to optimize the evacuation of irradiated fuel for reprocessing. The purpose of this paper is to describe these realization but as a preliminary, they will present briefly the importance of the transport of nuclear materials for EDF

PWR and BWR light water reactor systems in the USA and their fuel cycle

International Nuclear Information System (INIS)

Crawford, W.D.

1977-01-01

Light water reactor operating experience in the USA can be considered to date from the choice of the pressurized water reactor (PWR) for use in the naval reactor program and the subsequent construction and operation of the nuclear power plant at Shippingport, Pennsylvania in 1957. The development of the boiling water reactor (BWR) in 1954 and its selection for the plant at Dresden, Illinois in 1959 established this concept as the other major reactor type in the US nuclear power program. The subsequent growth profile is presented, leading to 31 PWR's and 23 BWR's currently in operation as well as to plants in the planning and construction phase. A significant operating record has been accumulated concerning the availability of each of these reactor types as determined by: (1) outage for refueling, (2) component reliability, (3) maintenance requirements, and (4) retrofitting required by government regulation. In addition, the use and performance of BWR's and PWR's in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to assure effective safeguards at nuclear power installations. Safeguards measures currently in place are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring to verify that results are within the limits established in the licensing process. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. The PWR and BWR Fuel Cycle is examined in terms of: (1) fuel burnup experience and prospects for improvement, (2) the status and outlook for natural uranium resources, (3) enrichment capacity, (4) reprocessing and recycle, and the interrelationships among the latter three factors. High level waste management currently involving on-site storage of spent fuel is discussed

Criticality analysis for mixed thorium-uranium fuel in the Angra-2 PWR reactor using KENO-VI

Energy Technology Data Exchange (ETDEWEB)

Wichrowski, Caio C.; Gonçalves, Isadora C.; Oliveira, Claudio L.; Vellozo, Sergio O.; Baptista, Camila O., E-mail: wichrowski@ime.eb.br, E-mail: isadora.goncalves@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

2017-07-01

The increasing energy demand associated to the current sustainability challenges have given the thorium nuclear fuel cycle renewed interest in the scientific community. Studies have focused on energy production in different reactor designs through the fission of uranium 233, the product of thorium fertilization by neutrons. In order to make it possible for near future applications a strategy based on the adaptation of current nuclear reactors for the use of thorium fuels is being considered. In this work, bearing in mind these limitations, a code was used to evaluate the effect on criticality (k{sub inf}) of the mixing of thorium and uranium in different proportions in the fuel of a PWR, the German designed Angra-2 Brazilian reactor in order to scrutinise its behaviour and determine the feasibility of an adapted ThO{sub 2}-UO{sub 2} mixed fuel cycle using current PWR technology. The analysis is performed using the KENO-VI module in the SCALE 6.1 nuclear safety analysis simulation code and the information is taken from the Angra-2 FSAR (Final Security Analysis Report). (author)

Demonstration of uncertainty quantification and sensitivity analysis for PWR fuel performance with BISON

International Nuclear Information System (INIS)

Zhang, Hongbin; Zhao, Haihua; Zou, Ling; Burns, Douglas; Ladd, Jacob

2017-01-01

BISON is an advanced fuels performance code being developed at Idaho National Laboratory and is the code of choice for fuels performance by the U.S. Department of Energy (DOE)’s Consortium for Advanced Simulation of Light Water Reactors (CASL) Program. An approach to uncertainty quantification and sensitivity analysis with BISON was developed and a new toolkit was created. A PWR fuel rod model was developed and simulated by BISON, and uncertainty quantification and sensitivity analysis were performed with eighteen uncertain input parameters. The maximum fuel temperature and gap conductance were selected as the figures of merit (FOM). Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis. (author)

Demonstration of Uncertainty Quantification and Sensitivity Analysis for PWR Fuel Performance with BISON

Energy Technology Data Exchange (ETDEWEB)

Zhang, Hongbin; Ladd, Jacob; Zhao, Haihua; Zou, Ling; Burns, Douglas

2015-11-01

BISON is an advanced fuels performance code being developed at Idaho National Laboratory and is the code of choice for fuels performance by the U.S. Department of Energy (DOE)’s Consortium for Advanced Simulation of Light Water Reactors (CASL) Program. An approach to uncertainty quantification and sensitivity analysis with BISON was developed and a new toolkit was created. A PWR fuel rod model was developed and simulated by BISON, and uncertainty quantification and sensitivity analysis were performed with eighteen uncertain input parameters. The maximum fuel temperature and gap conductance were selected as the figures of merit (FOM). Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis.

A Critical Review of Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage

International Nuclear Information System (INIS)

Wagner, J.C.; Parks, C.V.

2000-01-01

This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k inf estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k inf estimates using the actual spent fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. This observation is considered to be important, especially considering the recent allowance of credit for soluble boron up to 5% in reactivity. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE for different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity. A significant concentration (approx. 2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion (le 500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of the REFFE

Modifications needed to operate PWR's plants in G-Mode

International Nuclear Information System (INIS)

Stainman, J.P.

1985-01-01

The production of electricity from PWR nuclear plants represents 44% of the total production of electricity in France for 1984, and 68% of the electricity produced by Thermal power plants (127 TWh over 187 TWh). These data show clearly that the French PWR plants do not work in ''base mode'' anymore but have to fit production with consumption, in other words to assume the frequency control. To participate permanently to the load follow and frequency control, it appeared that some improvements in the field of pressurizer level and pressure control were necessary as well as in the field of operator aids computer. It should be noted that these improvements are useful even without taking into account the constraints due to load follow and frequency control because of the mechanical stress in the CVCS piping, for instance. Some additional tests are planned to better identify this specific problem. The need of a more flexible operating mode than ones given by the initial system (black control rods), significantly reduced in 1973 due to the application of the ECCS criterion, led EDF and Framatome to develop a new operating mode (G. Mode) allowing a faster power escalation (5% PN/mn) whatever the fuel burn-up. This new operating mode improves significantly also the flexibility of operation when the frequency control is needed, and helps a lot the operators in such cases. All the 900 MWe Nuclear plants will be able to operate in ''G mode'' before the end of 1984

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

2003 annual results of EdF group; Resultats annuels 2003 du groupe EDF

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-03-01

The Electricite de France (EdF) group Board of Directors, meeting on March 11, 2004, under the Chairmanship of Francois Roussely, reviewed the audited consolidated financial statements for the year ended 12/31/2003. This document presents the consolidated results of EdF group for 2003: consolidated financial statements, highlights of the year, focus on 2003 events, commercial results, EDF in Europe and worldwide, EDF France highlights, key figures. (J.S.)

In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

International Nuclear Information System (INIS)

Gunnerson, F.S.; MacDonald, P.E.

1980-01-01

The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

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

International Nuclear Information System (INIS)

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

1991-11-01

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

A Hold-down Margin Assessment using Statistical Method for the PWR Fuel Assembly

International Nuclear Information System (INIS)

Jeon, S. Y.; Park, N. K.; Lee, K. S.; Kim, H. K.

2007-01-01

The hold-down springs provide an acceptable hold down force against hydraulic uplift force absorbing the length change of the fuel assembly relative to the space between the upper and lower core plates in PWR. These length changes are mainly due to the thermal expansion, irradiation growth and creep down of the fuel assemblies. There are two kinds of hold-down springs depending on the different design concept of the reactor internals of the PWR in Korea, one is a leaf-type hold down spring for Westinghouse type plants and the other is a coil-type hold-down spring for OPR1000 (Optimized Power Reactor 1000). There are four sets of hold-down springs in each fuel assembly for leaf type hold-down spring and each set of the hold-down springs consists of multiple tapered leaves to form a cantilever leaf spring set. The length, width and thickness of the spring leaves are selected to provide the desired spring constant, deflection range, and hold down force. There are four coil springs in each fuel assembly for coil-type hold-down spring. In this study, the hold-down forces and margins were calculated for the leaf-type and coil-type hold-down springs considering geometrical data of the fuel assembly and its components, length changes of the fuel assembly due to thermal expansion, irradiation growth, creep, and irradiation relaxation. The hold-down spring forces were calculated deterministically and statistically to investigate the benefit of the statistical calculation method in view of hold-down margin. The Monte-Carlo simulation method was used for the statistical hold down force calculation

EDF, the 2003-2007 project; EDF, le projet 2003-2007

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-07-01

This brochure presents the 2003-2007 project of Electricite de France (EdF) in the framework of the opening of European energy markets. The aim of this project is to take over the present day contract of EdF with the French government in order to become an anonymous company. The wish of F. Roussely, president of EdF, is to open and increase the capital of the company but also to fulfill its public utility missions. The stake of EdF is to stay competitive, to become a European energy leader, to stay an integrated company combining production, transport, distribution and commercialization, and finally to maintain its historical public utility standards: specific tariff conditions and services for low income people, financing of renewable energy projects. (J.S.)

Thermal behaviour of high burnup PWR fuel under different fill gas conditions

International Nuclear Information System (INIS)

Tverberg, T.

2001-01-01

During its more than 40 years of existence, a large number of experiments have been carried out at the Halden Reactor Project focusing on different aspects related to nuclear reactor fuel. During recent years, the fuels testing program has mainly been focusing on aspects related to high burnup, in particular in terms of fuel thermal performance and fission gas release, and often involving reinstrumentation of commercially irradiated fuel. The paper describes such an experiment where a PWR rod, previously irradiated in a commercial reactor to a burnup of ∼50 MWd/kgUO 2 , was reinstrumented with a fuel central oxide thermocouple and a cladding extensometer together with a high pressure gas flow line, allowing for different fill gas compositions and pressures to be applied. The paper focuses on the thermal behaviour of such LWR rods with emphasis on how different fill gas conditions influence the fuel temperatures and gap conductance. Rod growth rate was also monitored during the irradiation in the Halden reactor. (author)

Analysis of differences in fuel safety criteria for WWER and western PWR nuclear power plants

International Nuclear Information System (INIS)

2003-11-01

In 2001 the OECD issued a report of the NEA/CSNI (Committee on the Safety of Nuclear Installations) Task Force on the existing safety criteria for reactor fuel for western LWR nuclear power plants (both for PWRs and BWRs) under new design elements. Likewise in 2001, the IAEA released a report by a Working Group on the existing safety criteria for reactor fuel for WWER nuclear power plants under new design requirements. However, it was found that it was not possible to compare the two sets of criteria on the basis upon which they had been established. Therefore, the IAEA initiated an assessment of the common features and differences in fuel safety criteria between plants of eastern and western design, focusing on western PWRs and eastern WWER reactors. Between October 2000 and November 2001, the IAEA organized several workshops with representatives from eastern and western European countries in which the current fuel safety related criteria for PWR and WWER reactors were reviewed and compared. The workshops brought together expert representatives from the Russian Federation, from the Ukraine and from western countries that operate PWRs. The first workshop focused on a general overview of the fuel safety criteria in order for all representatives to appreciate the various criteria and their respective bases. The second workshop (which involved one western and one eastern expert) concentrated on addressing and explaining the differences observed, and documenting all these results in preparation for a panel discussion. This panel discussion took place during the third workshop, where the previously obtained results were reviewed in detail and final recommendations were made. This report documents the findings of the workshops. It highlights the common features and differences between PWR and WWER fuel, and may serve as a general basis for the safety evaluation of these fuels. Therefore, it will be very beneficial for licensing activities for PWR and WWER plants, as it

Minor actinide transmutation on PWR burnable poison rods

International Nuclear Information System (INIS)

Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

2015-01-01

Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

EDF at a glance 2003; Groupe EDF - organisation et chiffres cles 2003

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

This document presents the organization and key data of Electricite de France (EdF) group for 2003: business, priorities, values, different branches of the group (continental Europe, western Europe, Mediterranean branch, Africa, Asia-Pacific, Americas, Dalkia energy services, development branch), management, board of directors, market opening in France, human resource strategy, R and D, customers (residential, commercial, business and industrial, local authorities), industrial tool (fossil-fired generation, network management, renewable energy sources), reliability, financial flexibility, sustainable development (role of renewables in generation, CO{sub 2} emissions, radiation protection, nuclear waste, environment, handling fuel poverty, landscape conservation), contacts and affiliates outside France. (J.S.)

Thermal hydraulic design of a hydride-fueled inverted PWR core

International Nuclear Information System (INIS)

Malen, J.A.; Todreas, N.E.; Hejzlar, P.; Ferroni, P.; Bergles, A.

2009-01-01

An inverted PWR core design utilizing U(45%, w/o)ZrH 1.6 fuel (here referred to as U-ZrH 1.6 ) is proposed and its thermal hydraulic performance is compared to that of a standard rod bundle core design also fueled with U-ZrH 1.6 . The inverted design features circular cooling channels surrounded by prisms of fuel. Hence the relative position of coolant and fuel is inverted with respect to the standard rod bundle design. Inverted core designs with and without twisted tape inserts, used to enhance critical heat flux, were analyzed. It was found that higher power and longer cycle length can be concurrently achieved by the inverted core with twisted tape relative to the optimal standard core, provided that higher core pressure drop can be accommodated. The optimal power of the inverted design with twisted tape is 6869 MW t , which is 135% of the optimally powered standard design (5080 MW t -determined herein). Uncertainties in this design regarding fuel and clad dimensions needed to accommodate mechanical loads and fuel swelling are presented. If mechanical and neutronic feasibility of these designs can be confirmed, these thermal assessments imply significant economic advantages for inverted core designs.

Effects of fabrication requirements on fuel performance in relation to operating conditions. The views of Electricite de France (EDF)

International Nuclear Information System (INIS)

Ponticq, M.; Richer, P.; Scribe, G.

1979-01-01

Because of the operational constraints relating to fuel behaviour, imperfect knowledge of the behaviour of a defective fuel assembly and, in the near future, the need to adapt reactor power to grid following (load following and remote control), EDF is aiming to reduce the present rate of fuel failure. While the phenomena affecting fuel behaviour have now been listed and analysed, the efforts at reducing their consequences have yet to be completed. This can be achieved, firstly, by reducing or eliminating fabrication defects, which are responsible for failure at the beginning of fuel life, through establishment of a good quality assurance organization and the search for still higher efficiency of quality control and fabrication equipment, and secondly, by developing fabrication techniques minimizing in particular cladding-pellet interactions and the stress corrosion of the cladding, which are responsible for fuel failure as from mid-life. However, the reactor operating conditions likely to apply in the near future may lead, for a given fuel configuration, to a re-evaluation of the fabrication parameters of cladding and UO 2 pellets. (author)

The computer program ELCOM in the planning and structural analysis of PWR fuel elements: an example

International Nuclear Information System (INIS)

Silva Macedo, L.V. da

1990-01-01

Is's presented some results obtained with the ELCOM computer code, such as deflections, moments and natural frequencies, used in the design and structural analysis of PWR fuels assemblies. It's studied the behavior of these results varying the number of spacer grids, the rigidity of the joint between the fuel pin and the spacer grid, and the fuel assembly's boundary condition, considered in the analysis, in it's mounting into the core (if clamped-clamped, clamped-hinged or hinged-hinged). (author)

Introduction of advanced pressurized water reactors in France

International Nuclear Information System (INIS)

Millot, J.P.; Nigon, M.; Vitton, M.

1988-01-01

Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

Developing and analyzing long-term fuel management strategies for an advanced Small Modular PWR

Energy Technology Data Exchange (ETDEWEB)

Hedayat, Afshin, E-mail: ahedayat@aeoi.org.ir

2017-03-15

Highlights: • Comprehensive introduction and supplementary concepts as a review paper. • Developing an integrated long-term fuel management strategy for a SMR. • High reliable 3-D core modeling over fuel pins against the traditional LRM. • Verifying the expert rules of large PWRs for an advanced small PWR. • Investigating large numbers of safety parameters coherently. - Abstract: In this paper, long-term fuel management (FM) strategies are introduced and analyzed for a new advanced Pressurized Light Water Reactor (PWR) type of Small Modular Reactors (SMRs). The FM strategies are developed to be safe and practical for implementation as much as possible. Safety performances, economy of fuel, and Quality Assurance (QA) of periodic equilibrium conditions are chosen as the main goals. Flattening power density distribution over fuel pins is the major method to ensure safety performance; also maximum energy output or permissible discharging burn up indicates economy of fuel fabrication costs. Burn up effects from BOC to EOC have been traced, studied, and highly visualized in both of transport lattice cell calculations and diffusion core calculations. Long-term characteristics are searched to gain periodical equilibrium characteristics. They are fissile changes, neutron spectrum, refueling pattern, fuel cycle length, core excess reactivity, average, and maximum burn up of discharged fuels, radial Power Peaking Factors (PPF), total PPF, radial and axial power distributions, batch effects, and enrichment effects for fine regulations. Traditional linear reactivity model have been successfully simulated and adapted via fine core and burn up calculations. Effects of high burnable neutron poison and soluble boron are analyzed. Different numbers of batches via different refueling patterns have been studied and visualized. Expert rules for large type PWRs have been influenced and well tested throughout accurate equilibrium core calculations.

2002 annual report EDF group; 2002 rapport annuel groupe EDF

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-07-01

This document is the 2002 annual report of Electricite de France (EdF) group, the French electric utility. Content: Introductory section (EDF at a glance, Chairman's message, 2002 Highlights); Corporate governance and Group strategy (Corporate governance, sustainable growth strategy, EDF branches); Financial performance (Reaching critical mass, Margins holding up well, Balance sheet); Human resources (Launching Group-wide synergies, Optimising human resources); Customers (Major customers, SMEs and professional customers, Local authorities, Residential customers, Ensuring quality access to electricity); Generation (A balanced energy mix, Nuclear generation, Fossil-fuelled generation, Renewable energies); Corporate social responsibility (Global and local partnerships, Promoting community development)

2002 annual report EDF group; 2002 rapport annuel groupe EDF

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-07-01

This document is the 2002 annual report of Electricite de France (EdF) group, the French electric utility. Content: Introductory section (EDF at a glance, Chairman's message, 2002 Highlights); Corporate governance and Group strategy (Corporate governance, sustainable growth strategy, EDF branches); Financial performance (Reaching critical mass, Margins holding up well, Balance sheet); Human resources (Launching Group-wide synergies, Optimising human resources); Customers (Major customers, SMEs and professional customers, Local authorities, Residential customers, Ensuring quality access to electricity); Generation (A balanced energy mix, Nuclear generation, Fossil-fuelled generation, Renewable energies); Corporate social responsibility (Global and local partnerships, Promoting community development)

Determination of burnup, cooling time and initial enrichment of PWR spent fuel by use of gamma-ray activity ratios

International Nuclear Information System (INIS)

Min, D.K.; Park, H.J.; Park, K.J.; Ro, S.G.; Park, H.S.

1999-01-01

The Korea Atomic Energy Institute has been developing the algorithms for sequential determination of cooling time, initial enrichment and burnup of the PWR spent fuel assembly by use of gamma ratio measurements, i.e. 134 Cs/ 137 Cs, 154 Eu/ 137 Cs and 106 Ru 137 Cs/( 134 Cs) 2 . Calculations were performed by applying the ORIGEN-S code. This method has advantages over combination techniques of neutron and gamma measurement, because of its simplicity and insensitivity to the measurement geometry. For verifying the algorithms an experiment for determining the cooling time, initial enrichment and burnup of the two PWR spent fuel rods was conducted by use of high-resolution gamma detector (HPGe) system only. This paper describes the method used and interim results of the experiment. This method can be applied for spent fuel characterization, burnup credit and safeguards of the spent fuel management facility

SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

Energy Technology Data Exchange (ETDEWEB)

Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL

2010-03-01

The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

Simulation of nonlinear dynamics of a PWR core by an improved lumped formulation for fuel heat transfer

International Nuclear Information System (INIS)

Su, Jian; Cotta, Renato M.

2000-01-01

In this work, thermohydraulic behaviour of PWR, during reactivity insertion and partial loss-of-flow, is simulated by using a simplified mathematical model of reactor core and primary coolant. An improved lumped parameter formulation for transient heat conduction in fuel rod is used for core heat transfer modelling. Transient temperature response of fuel, cladding and coolant is analysed. (author)

Burn-up credit in criticality safety of PWR spent fuel

Energy Technology Data Exchange (ETDEWEB)

Mahmoud, Rowayda F., E-mail: Rowayda_mahmoud@yahoo.com [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)

2014-12-15

Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

EDF's experience with supercomputing and challenges ahead - towards multi-physics and multi-scale approaches

Energy Technology Data Exchange (ETDEWEB)

Delbecq, J.M.; Banner, D. [Electricite de France (EDF)- R and D Division, 92 - Clamart (France)

2003-07-01

Nuclear power plants are a major asset of the EDF company. To remain so, in particular in a context of deregulation, competitiveness, safety and public acceptance are three conditions. These stakes apply both to existing plants and to future reactors. The purpose of the presentation is to explain how supercomputing can help EDF to satisfy these requirements. Three examples are described in detail: ensuring optimal use of nuclear fuel under wholly safe conditions, understanding and simulating the material deterioration mechanisms and moving forward with numerical simulation for the performance of EDF's activities. In conclusion, a broader vision of EDF long term R and D in the field of numerical simulation is given and especially of five challenges taken up by EDF together with its industrial and scientific partners. (author)

Proposal for a advanced PWR core with adequate characteristics for passive safety concept

International Nuclear Information System (INIS)

Perrotta, Jose Augusto

1999-01-01

This work presents a discussion upon the suitable from an advanced PWR core, classified by the EPRI as 'Passive PWR' (advanced reactor with passive safety concept to power plants with less than 600 MW electrical power). The discussion upon the type of core is based on nuclear fuel engineering concepts. Discussion is made on type of fuel materials, structural materials, geometric shapes and manufacturing process that are suitable to produce fuel assemblies which give good performance for this type of reactors. The analysis is guided by the EPRI requirements for Advanced Light Water Reactor (ALWR). By means of comparison, the analysis were done to Angra 1 (old type of 600 MWe PWR class), and the design of the Westinghouse Advanced PWR-AP600. It was verified as a conclusion of this work that the modern PWR fuels are suitable for advanced PWR's Nevertheless, this work presents a technical alternative to this kind of fuel, still using UO 2 as fuel, but changing its cylindrical form of pellets and pin type fuel element to plane shape pallets and plate type fuel element. This is not a novelty fuel, since it was used in the 50's at Shippingport Reactor and as an advanced version by CEA of France in the 70's. In this work it is proposed a new mechanical assembly design for this fuel, which can give adequate safety and operational performance to the core of a 'Passive PWR'. (author)

Qualification of the testing process by a new rotating probe for the 900 MW PWR steam generators tubes

International Nuclear Information System (INIS)

Caston, D.

2001-01-01

In 1997, EDF invites bids for the development of eddy current probes, in order to better control the steam generator tubes rolling zone of the PWR. After examination of the response, EDF accepted a prototype of a rotating probe. The EDF specifications of the new steam generators contract, fix the inspection performance and rhythm of the rotating probe, which implements two different technic: the STL (long rotating prob)e for the detection and characterization of longitudinal defects; the STT (transverse rotating probe) for the detection and characterization of circumferential defects. The new rotating probe should not increase the control time. Two new equipments have been developed to implement this new probe: a new support and a software of online inspection with data processing and analyzing. Abstract only. (A.L.B.)

Neutron multiplication and shielding problems in PWR spent-fuel shipping casks

International Nuclear Information System (INIS)

Devillers, C.

1976-01-01

In order to evaluate the degree of accuracy of computational methods used for the shield design of spent-fuel shipping casks, comparisons were made between biological dose rate calculations and measurements at the surface of a cask carrying three PWR fuel assemblies (the fuel being successively wet and dry). The experimental methods used provide ksub(eff) with an accuracy of 0.024. Neutron multiplication coefficients provided by the APOLLO and DOT-3 codes are located within the uncertainty range of the experimentally derived values. The APOLLO plus DOT codes for neutron source calculations and ANISN plus DOT codes for neutron transmission calculations provide neutron dose rate predictions in agreement with measurements to within 10%. The PEPIN 76 code used for deriving fission product γ-rays and the point kernel code MERCURE 4 treating the γ-ray transmission give γ dose rate predictions that generally differ from measurements by less than 25%

Analysis of confinement effects for in-water seismic tests on PWR fuel assemblies

International Nuclear Information System (INIS)

Broc, Daniel; Queval, Jean-Claude; Rigaudeau, J.; Viallet, E.

2001-01-01

In the framework of a comprehensive program on the seismic behaviour of the PWR reactor cores, tests have been performed on a row of six PWR fuel assemblies, with two confinement configurations in water. Global fluid motion along the row is not allowed in the 'full confinement configuration', and is allowed in the 'lateral confinement configuration'. The seismic test results show that the impact forces at assembly grid levels are significantly smaller with the full confinement. This is due to damping, which is found to be larger in this configuration where the average fluid velocity inside the assembly (around the rods) is itself larger. We present analyses of these phenomena from theoretical and experimental standpoint. This involves both fluid models and structural models of the assembly row. (author)

Data assimilation and PWR primary measurement

International Nuclear Information System (INIS)

Mercier, Thibaud

2015-01-01

A Pressurized Water Reactor (PWR) Reactor Coolant System (RCS) is a highly complex physical process: heterogeneous power, flow and temperature distributions are difficult to be accurately measured, since instrumentations are limited in number, thus leading to the relevant safety and protection margins. EDF R and D is seeking to assess the potential benefits of applying Data Assimilation to a PWR's RCS (Reactor Coolant System) measurements, in order to improve the estimators for parameters of a reactor's operating setpoint, i.e. improving accuracy and reducing uncertainties and biases of measured RCS parameters. In this thesis, we define a 0D semi-empirical model for RCS, satisfying the description level usually chosen by plant operators, and construct a Monte-Carlo Method (inspired from Ensemble Methods) in order to use this model with Data Assimilation tools. We apply this method on simulated data in order to assess the reduction of uncertainties on key parameters: results are beyond expectations, however strong hypothesis are required, implying a careful preprocessing of input data. (author)

Monte Carlo Simulation of Quantitative Electron Probe Microanalysis of the PWR Spent Fuel with a Pt Coating

International Nuclear Information System (INIS)

Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum

2012-01-01

The PWR spent fuel sample should be coated with conducting material in order to provide a path for electrons and to prevent charging. Generally, the ZAF method has been used for quantitative electron probe microanalysis of conducting samples. However, the coated samples are not applicable for the ZAF method. Probe current, primary electron energy and x-ray produced by the primary beam are attenuated within the coating films. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program [2] to evaluate the x-ray attenuation within the Pt coating films. The target samples are the PWR spent fuels with 50 GWd/tU of burnup , 6 years of cooling time and a Pt coating film (3, 5, 7, 10 and 15 nm thickness)

Monte Carlo Simulation of Quantitative Electron Probe Microanalysis of the PWR Spent Fuel with a Pt Coating

Energy Technology Data Exchange (ETDEWEB)

Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

The PWR spent fuel sample should be coated with conducting material in order to provide a path for electrons and to prevent charging. Generally, the ZAF method has been used for quantitative electron probe microanalysis of conducting samples. However, the coated samples are not applicable for the ZAF method. Probe current, primary electron energy and x-ray produced by the primary beam are attenuated within the coating films. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program [2] to evaluate the x-ray attenuation within the Pt coating films. The target samples are the PWR spent fuels with 50 GWd/tU of burnup , 6 years of cooling time and a Pt coating film (3, 5, 7, 10 and 15 nm thickness)

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Development of a reference spent fuel library of 17x17 PWR fuel assemblies

International Nuclear Information System (INIS)

Rossa, Riccardo; Borella, Alessandro; Van der Meer, Klaas

2013-01-01

One of the most common ways to investigate new Non-Destructive Assays (NDA) for the spent fuel assemblies are Monte Carlo simulations. In order to build realistic models the user must define in an accurate way the material compositions and the source terms in the system. This information can be obtained using burnup codes such as ORIGEN-ARP and ALEPH2.2, developed at SCK-CEN. These software applications allow the user to select the irradiation history of the fuel assembly and to calculate the corresponding isotopic composition and neutron/gamma emissions as a function of time. In the framework of the development of an innovative NDA for spent fuel verifications, SCK•CEN built an extensive fuel library for 17x17 PWR assemblies, using both ORIGEN-ARP and ALEPH2.2. The parameters considered in the calculations were initial enrichment, discharge burnup, and cooling time. The combination of these variables allows to obtain more than 1500 test cases. Considering the broad range of the parameters, the fuel library can be used for other purposes apart from spent fuel verifications, for instance for the direct disposal in geological repositories. In addition to the isotopic composition of the spent fuel, the neutron and photon emissions were also calculated and compared between the two codes. The comparison of the isotopic composition showed a good agreement between the codes for most of the relevant isotopes in the spent fuel. However, specific isotopes as well as neutron and gamma spectra still need to be investigated in detail.

Re-irradiation and limit testing of the fuels PWR type reactors

International Nuclear Information System (INIS)

Roche, M.; Molvault, M.

1978-01-01

In view of investigating the neutron radiation behavior of PWR fuel pins, the S.P.S. (Services des Piles de Saclay) developed a set of experimental means used at OSIRIS in Saclay Nuclear Research Center. Said devices are shown to be able to meet present problems concerning can failures, power and temperature cycling, remote-control studies. These means can also be used either for statistical studies, they can then receive several samples, or for analytical studies in instrumented devices of large capacity and accelerated irradiation rate [fr

Development of an innovative PWR for low cost fuel recycle and waste reduction

International Nuclear Information System (INIS)

Kanagawa, Takashi; Onoue, Masaaki

2001-01-01

In order to bear long-term and stable energy supply, it is important for nuclear power generation to realize establishment of energy security controlling dependence on natural resources and reduction of long-life radioactive wastes such as minor actinide elements (MA) and so on. For this, establishment of fast breeder reproducible on its fuel and of fuel recycling is essential and construction of the fuel recycling capable of repeatedly recycling of plutonium (Pu) and MA with low cost is required. Here were proposed a fuel recycling system combining recycling type PWR with advanced recycling system under development for Na cooling fast breeder reactor as a candidate filling such conditions, to show its characteristics and effects after its introduction. By this system, some facilities to realize flexible and low cost fuel recycling, to reduce longer-life radioactive wastes due to recycling burning of Pu and MA, and to realize an electric power supplying system independent on natural resources due to fuel breeding feature, were shown. (G.K.)

Assessment of the insertion of reprocessed fuel spiked with thorium in a PWR core

Energy Technology Data Exchange (ETDEWEB)

Castro, Victor F.; Monteiro, Fabiana B.A.; Pereira, Claubia, E-mail: victorfc@fis.grad.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

Reprocessed fuel by UREX+ technique and spiked with thorium was inserted in a PWR core and neutronic parameters have been analyzed. Based on the Final Safety Analysis Report (FSAR) of the Angra-2 reactor, the core was modeled and simulated with SCALE6.0 package. The neutronic data evaluation was carried out by the analysis of the effective and infinite multiplication factors, and the fuel evolution during the burnup. The conversion ratio (CR) was also evaluated. The results show that, when inserting reprocessed fuel spiked with thorium, the insertion of burnable poison rods is not necessary, due to the amount of absorber isotopes present in the fuel. Besides, the conversion ratio obtained was greater than the presented by standard UO{sub 2} fuel, indicating the possibility of extending the burnup. (author)

New EdF's three challenges

International Nuclear Information System (INIS)

Lepetit, V.

2005-01-01

The opening of 15% of Electricite de France (EdF) capital to the private sector will allow the first world electric utility to raise 7 billions of euros of cash flow. EdF will have also to face the new European situation: a strained market, several consolidations to come and many competitors to deal with. EdF has to re-launch its production tool and to reinforce its interconnection capacities if it wants to stay the number one of power exports in Europe. The integration of Edison company (Italy) by EdF gives access to new gas capacities but other resources have to be found if EdF wants to enlarge its gas offer. A new concentration era is foreseen in Europe and EdF will have to seize the economic growth opportunities in particular in Europe and China. (J.S.)

Fuel rod D07/B15 from Ringhals 2 PWR: Source material for corrosion/leach tests in groundwater. Fuel rod/pellet characterization program. Pt. 1

International Nuclear Information System (INIS)

Forsyth, R.

1987-03-01

A joint SKB/STUDSVIK experimental program to determine the corrosion rates and to establish the corrosion mechanisms of spent UO 2 fuel in groundwater under both oxidizing and reducing conditions is in progress in the Hot Cell Laboratory of Studsvik Energiteknik AB. High burnup fuel of both BWR and PWR type are studied. Characterization of the spent fuel at both rod and pellet level is an important part of the experimental program. Experiments on PWR fuel have been concentrated so far on specimens from one rod, manufacturer's number 03688, which had occupied position B15 in assembly D07. This assembly had been irradiated for 5 cycles in the Ringhals 2 reactor between 1977 and 1983. The calculated assembly burnup was 41.3 MWd/kg U. The present report is a collection of separate reports describing those items in the characterization program which have been performed so far. No overall summary of the experimental results is given here, and the report should be viewed as a collection of reference data. (orig.)

Chemical and radiochemical specifications - PWR power plants; Specifications chimiques et radiochimiques - Centrales REP

Energy Technology Data Exchange (ETDEWEB)

Stutzmann, A [Electricite de France (EDF), 93 - Saint-Denis (France)

1997-07-01

Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

PWR fuel of high enrichment with erbia and enriched gadolinia

International Nuclear Information System (INIS)

Bejmer, Klaes-Håkan; Malm, Christian

2011-01-01

Today standard PWR fuel is licensed for operation up to 65-70 MWd/kgU, which in most cases corresponds to an enrichment of more than 5 w/o "2"3"5U. Due to criticality safety reason of storage and transportation, only fuel up to 5 w/o "2"3"5U enrichment is so far used. New fuel storage installations and transportation casks are necessary investments before the reactivity level of the fresh fuel can be significantly increased. These investments and corresponding licensing work takes time, and in the meantime a solution that requires burnable poisons in all pellets of the fresh high-enriched fuel might be used. By using very small amounts of burnable absorber in every pellet the initial reactivity can be reduced to today's levels. This study presents core calculations with fuel assemblies enriched to almost 6 w/o "2"3"5U mixed with a small amount of erbia. Some of the assemblies also contain gadolinia. The results are compared to a reference case containing assemblies with 4.95 w/o "2"3"5U without erbia, utilizing only gadolinia as burnable poison. The comparison shows that the number of fresh fuel assemblies can be reduced by 21% (which increases the batch burnup by 24%) by utilizing the erbia fuel concept. However, increased cost of uranium due to higher enrichment is not fully compensated for by the cost gain due to the reduction of the number assemblies. Hence, the fuel cycle cost becomes slightly higher for the high enrichment erbia case than for the reference case. (author)

Modeling of the PWR fuel mechanical behaviour and particularly study of the pellet-cladding interaction in a fuel rod

International Nuclear Information System (INIS)

Hourdequin, N.

1995-05-01

In Pressurized Water Reactor (PWR) power plants, fuel cladding constitutes the first containment barrier against radioactive contamination. Computer codes, developed with the help of a large experimental knowledge, try to predict cladding failures which must be limited in order to maintain a maximal safety level. Until now, fuel rod design calculus with unidimensional codes were adequate to prevent cladding failures in standard PWR's operating conditions. But now, the need of nuclear power plant availability increases. That leads to more constraining operating condition in which cladding failures are strongly influenced by the fuel rod mechanical behaviour, mainly at high power level. Then, the pellet-cladding interaction (PCI) becomes important, and is characterized by local effects which description expects a multidimensional modelization. This is the aim of the TOUTATIS 2D-3D code, that this thesis contributes to develop. This code allows to predict non-axisymmetric behaviour too, as rod buckling which has been observed in some irradiation experiments and identified with the help of TOUTATIS. By another way, PCI is influenced by under irradiation experiments and identified with the help of TOUTATIS which includes a densification model and a swelling model. The latter can only be used in standard operating conditions. However, the processing structure of this modulus provides the possibility to include any type of model corresponding with other operating conditions. In last, we show the result of these fuel volume variations on the cladding mechanical conditions. (author). 25 refs., 89 figs., 2 tabs., 12 photos., 5 appends

Instrumentation of fuel safety test rods of the PWR system in the Phebus reactor

International Nuclear Information System (INIS)

Schley, Robert; Leveque, J.P.; Aujollet, J.M.; Dutraive, Pierre; Colome, Jean; Bouly, J.C.

1979-01-01

The tests were performed in an experimental cell centred in the core of the PHEBUS water reactor of 50 MW. The CEA make two types of apparatus for testing the safety of PWR fuel. One is for testing a single fuel stick and the other a bunch of 25 sticks. The instrumentation described enables the main parameters of the test to be known: temperatures of the fuel - central temperature of the UO 2 - cladding surface temperatures; temperature of the cooling circuits - thermal balance - temperatures of the structures, etc.; coolant pressure; internal pressure of the fuel sticks; direction and flow rate of the fluid. This instrumentation and the technological problems to be overcome are described and the results of the first tests carried out are given [fr

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

Operating procedures for emergency situations in EDF PWR plants

International Nuclear Information System (INIS)

Depond, G.; Resse, L.

1992-01-01

Analysis of incidents and accidents occurring at French and foreign power plants - particularly the TMI accident - and the commissioning of many units in France, as well as tests on simulators, have all demonstrated that an improvement of safety in nuclear power units depends largely on the improvement of the man-machine interface and particularly of emergency operating procedures (EOP). EDF has taken numerous actions in this direction, especially since 1979. First of all, in improving the classical approach based on event-oriented procedures: Rewriting of initial accident operating procedures with regard to their technical contents their form, and the organization of the operating team (procedures I and A); Extension of initial procedures into areas at the limits of design basis and beyond the design basis limits (procedures H). Nevertheless, this approach is subject to several weaknesses. Dependence on a precise initial diagnosis, impossibility to take into account all the conceivable accidental situations, discrepancies between the predicted pattern and the reality. These drawbacks of the event approach have led us to revise the technical conception of the EOPs, and to develop a new approach based on a continuous monitoring of the physical states of the plant and the ability to define a relationship between the physical state of the plant and the operator actions. (author). 4 figs

Behaviour of fission products in PWR primary coolant and defected fuel rods evaluation

International Nuclear Information System (INIS)

Bourgeois, P.; Stora, J.P.

1979-01-01

The activity surveillance of the PWR primary coolant by γ spectometry gives some informations on fuel failures. The activity of different nuclides e.g. Xenons, Kryptons, Iodines, can be correlated with the number of the defected fuel rods. Therefore the precharacterization with eventually a prelocalization of the related fuel assemblies direct the sipping-test and allows a saving of time during refueling. A model is proposed to calculate the number of the defected rods from the activity measurements of the primary coolant. A semi-empirical model of the release of the fission products has been built from the activity measurements of the primary coolant in a 900 MWe PWR. This model allows to calculate the number of the defected rods and also a typical parameter of the mean damage. Fission product release is described by three stages: release from uranium dioxide, transport across the gas gap and behaviour in the primary coolant. The model of release from the oxide considers a diffusion process in the grains with trapping. The release then occurs either directly to free surfaces or with a delay due to a transit into closed porosity of the oxide. The amount released is the same for iodine and rare gas. With the gas gap transit is associated a transport time and a probability of trapping for the iodines. In the primary coolant the purification and the radioactive decay are considered. (orig.)

Road-map design for thorium-uranium breeding recycle in PWR - 031

International Nuclear Information System (INIS)

Shengyi, Si

2010-01-01

The paper was focused on designing a road-map to finally approach sustainable Thorium-Uranium ( 232 Th- 233 U) Breeding Recycle in current PWR, without any other change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. At first, the paper presented some insights to the inherence of Thorium-Uranium fuel conversion or breeding in PWR based on the neutronics theory and revealed the prerequisites for Thorium-Uranium fuel in PWR to achieve sustainable Breeding Recycle; And then, various Thorium-based fuels were designed and examined, and the calculation results further validated the above theoretical deductions; Based on the above theoretical analysis and calculation results, a road-map for sustainable Thorium-Uranium breeding recycle in PWR was outlined finally. (authors)

Mechanical behaviour of PWR fuel rods during intermediate storage

International Nuclear Information System (INIS)

Bouffioux, P.; Dalmas, R.; Bernaudat, C.

2000-01-01

EDF, which owns the irradiated fuel coming from its NPPs, has initiated studies regarding the mechanical behaviour of a fuel rod and the integrity of its cladding, in the case where the spent fuel is stored for a significant duration. During the phases following in-reactor irradiation (ageing in a water-pool, transport and intermediate storage), many phenomena, which are strongly coupled, may influence the cladding integrity: - residual power and temperature decay; - helium production and release in the free volume of the rod (especially for MOX fuel); - fuel column swelling; - cladding creep-out under the inner gas pressure of the fuel rod; - metallurgical changes due to high temperatures during transportation. In parallel, the quantification of the radiological risk is based on the definition of a cladding integrity criterion. Up to now, this criterion requires that the clad hoop strain due to creep-out does not exceed 1%. A more accurate criterion is being investigated. The study and modelling of all the phenomena mentioned above are included in a R and D programme. This programme also aims at redefining the cladding integrity criterion, which is assumed to be too conservative. The R and D programme will be presented. In order to predict the overall behaviour of the rod during the intermediate storage phases, the AVACYC code has been developed. It includes the models developed in the R and D programme. The input data of the AVACYC code are provided by the results of in-reactor rod behaviour simulations, using the thermal-mechanical CYRANO3 code. Its main results are the evolution vs. time of hoop stresses in the cladding, rod internal pressure and cladding hoop strains. Chained CYRANO-AVACYC calculations have been used to simulate the behaviour of MOX fuel rods irradiated up to 40 GWd/t and stored under air during 100 years, or under water during 50 years. For such fuels, where the residual power remains high, we show that a large part of the cladding strain

EDF'S ''OMF-structures'' RI-ISI process: overview of the methodology, applications and comparison with the other existing RI-ISI initiatives

International Nuclear Information System (INIS)

Codron, P.; Lauriot, B.; Dubreuil-Chambardel, A.

2001-01-01

The OMF-Structures process has been developed by EDF in order to optimise the maintenance of piping components. The principles of the OMF-Structures are consistent with the RCM concepts in EDF's OMF process (applied to active components). Several applications of this process have been performed. This paper also presents several applications which have been performed, notably on the AFW, RHR, and HPSI/LPSI systems of EDF's 900 MW PWR plants. These applications led to propose optimised preventive maintenance programs for these systems. These updated preventive maintenance programs confirmed and justified a large proportion of existing maintenance, despite some inspection tasks were suppressed and some inspection tasks were added. Globally, these new programs guarantee a better safety level, insuring that preventive maintenance is applied on piping segments where risks are concentrated. A comparative evaluation of the OMF-Structures process with the EPRI and WOG/ASME RI-ISI methodologies has been performed. This comparison demonstrated the consistence of the OMF-Structures principles and selection criteria with the other RI-ISI processes. (author)

The photovoltaic ambitious of EDF

International Nuclear Information System (INIS)

Houot, G.

2008-01-01

Added to the wind energy, EDF develops the photovoltaic by its subsidiaries EDF Energies Nouvelles, for the big power plants and EDF Energies Nouvelles Reparties centralized on the market of small installations for roofs. The author analyzes the society management and project concerning the photovoltaic development. (A.L.B.)

EDF Energies Nouvelles - 2010 Registration Document

International Nuclear Information System (INIS)

2011-01-01

EDF Energies Nouvelles is a world leader in renewable energy electricity. The company develops, builds and operates clean energy power plants both for its own account and for third parties. Historically, EDF Energies Nouvelles primarily developed its business in two geographical areas, Europe and North America (U.S., Canada and Mexico). EDF Energies Nouvelles is a subsidiary of EDF, helping the Group to achieve its renewable energy goals. The EDF Group generates low-carbon electricity around the world and actively participates in the energy transition. EDF Energies Nouvelles prioritizes development of wind and photovoltaic solar capacity. As an integrated operator with global reach, EDF Energies Nouvelles covers the entire renewable energy chain, from development to operation and maintenance, and manages all project phases in-house. This document is EDF Energies Nouvelles' registration document for the year 2010. It contains information about Group profile, governance, business, investments, property, plant and equipment, management, financial position, employees, shareholders, etc. The document includes the half-year and full year financial reports

Out-pile test of non-instrumented capsule for the advanced PWR fuel pellets in HANARO irradiation test

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Lee, C. B.; Oh, D. S.; Bang, J. K.; Kim, Y. M.; Yang, Y. S.; Jeong, Y. H.; Jeon, H. K.; Ryu, J. S. [KAERI, Taejon (Korea, Republic of)

2002-05-01

Non-instrumental capsule were designed and fabricated to irradiate the advanced pellet developed for the high burn-up LWR fuel in the HANARO in-pile capsule. This capsule was out-pie tested at Cold Test Loop-I in KAERI. From the pressure drop test results, it is noted that the flow velocity across the non-instrumented capsule of advanced PWR fuel pellet corresponding to the pressure drop of 200 kPa is measured to be about 7.45 kg/sec. Vibration frequency for the capsule ranges from 13.0 to 32.3 Hz. RMS displacement for non-instrumented capsule of advanced PWR fuel pellet is less than 11.6 {mu}m, and the maximum displacement is less that 30.5 {mu}m. The flow rate for endurance test were 8.19 kg/s, which was 110% of 7.45 kg/s. And the endurance test was carried out for 100 days and 17 hours. The test results found not to the wear satisfied to the limits of pressure drop, flow rate, vibration and wear in the non-instrumented capsule.

Balance and behavior of gaseous radionuclides released during initial PWR fuel reprocessing operations

International Nuclear Information System (INIS)

Leudet, A.; Miquel, P.; Goumondy, P.J.; Charrier, G.

1982-08-01

Five fuel pins, taken from a PWR fuel assembly with 32000 MWD/t burn-up were chopped and dissolved in leak-proof equipment designed for accurate determination of the composition and quantity of gaseous elements released in these operations. Analytical methods were specially developped to determine directly the noble gases, tritium and gaseous carbon compounds in the gas phase. Volatile iodine was kept as close as possible to the source by cold traps, then transferred to a caustic solution for quantitative analysis. The quantities and activities of gaseous fission products thus determined were compared with predicted values obtained through computation. Very good agreement was generally observed

Balance and behavior of gaseous radionuclides released during initial PWR fuel reprocessing operations

International Nuclear Information System (INIS)

Leudet, A.; Miquel, P.; Goumondy, P.J.; Charrier, G.

1983-01-01

Five fuel pins, taken from a PWR fuel assembly with 32,000 MwD/t burn-up were chopped and dissolved in leak-proof equipment designed for accurate determination of the composition and quantity of gaseous elements released in these operations. Analytical methods were specially developed to determine directly the noble gases, tritium and gaseous carbon compounds in the gas phase. Volatile iodine was kept as close as possible to the source by cold traps, then transferred to a caustic solution for quantitative analysis. The quantities and activities of gaseous fission products thus determined were compared with predicted values obtained through computation. Very good agreement was generally observed

The effects of fission gas release on PWR fuel rod design and performance

International Nuclear Information System (INIS)

Leech, W.J.; Kaiser, R.S.

1980-01-01

The purpose of this investigation was to determine the effects of fission gas release on PWR fuel rod design and performance. Empirical models were developed from fission gas release data. Fission gas release during normal operation is a function of burnup. There is little additional fission gas release during anticipated transients. The empirical models were used to evaluate Westinghouse fuel rod designs. It was determined that fission gas release is not a limiting parameter for obtaining rod average burnups in the range of 50,000 to 60,000 MWD/MTU. Fission gas release during anticipated transients has a negligible effect on the margins to rod design limits. (author)

Use of complex electronic equipment within radiative areas of PWR power plants: feability study

International Nuclear Information System (INIS)

Fremont, P.; Carquet, M.

1988-01-01

EDF has undertaken a study in order to evaluate the technical and economical feasibility of using complex electronic equipment within radiative areas of PWR power plants. This study lies on tests of VLSI components (Random Access Memories) under gamma rays irradiations, which aims are to evaluate the radiation dose that they can withstand and to develop a selection method. 125 rad/h and 16 rad/h tests results are given [fr

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

Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

International Nuclear Information System (INIS)

Wiles, L.E.; McCann, R.A.

1981-09-01

This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

Development and application of methods and computer codes of fuel management and nuclear design of reload cycles in PWR

International Nuclear Information System (INIS)

Ahnert, C.; Aragones, J.M.; Corella, M.R.; Esteban, A.; Martinez-Val, J.M.; Minguez, E.; Perlado, J.M.; Pena, J.; Matias, E. de; Llorente, A.; Navascues, J.; Serrano, J.

1976-01-01

Description of methods and computer codes for Fuel Management and Nuclear Design of Reload Cycles in PWR, developed at JEN by adaptation of previous codes (LEOPARD, NUTRIX, CITATION, FUELCOST) and implementation of original codes (TEMP, SOTHIS, CICLON, NUDO, MELON, ROLLO, LIBRA, PENELOPE) and their application to the project of Management and Design of Reload Cycles of a 510 Mwt PWR, including comparison with results of experimental operation and other calculations for validation of methods. (author) [es

Preliminary conceptual design of a geological disposal system for high-level wastes from the pyroprocessing of PWR spent fuels

Energy Technology Data Exchange (ETDEWEB)

Choi, Heui-Joo, E-mail: hjchoi@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daeduk-Daero, Yuseong, Daejon 305-353 (Korea, Republic of); Lee, Minsoo; Lee, Jong Youl [Korea Atomic Energy Research Institute, 1045 Daeduk-Daero, Yuseong, Daejon 305-353 (Korea, Republic of)

2011-08-15

Highlights: > A geological disposal system consists of disposal overpacks, a buffer, and a deposition hole or a disposal tunnel for high-level wastes from a pyroprocessing of PWR spent fuels is proposed. The amount and characteristics of high-level wastes are analyzed based on the material balance of pyroprocessing. > Four kinds of deposition methods, two horizontal and two vertical, are proposed. Thermal design is carried out with ABAQUS program. The spacing between the disposal modules is determined for the peak temperature in buffer not to exceed 100 deg. C. > The effect of the double-layered buffer is compared with the traditional single-layered buffer in terms of disposal density. Also, the effect of cooling time (aging) is illustrated. > All the thermal calculations are represented by comparing the disposal area of PWR spent fuels with the same cooling time. - Abstract: The inventories of spent fuels are linearly dependent on the production of electricity generated by nuclear energy. Pyroprocessing of PWR spent fuels is one of promising technologies which can reduce the volume of spent fuels remarkably. The properties of high-level wastes from the pyroprocessing are totally different from those of spent fuels. A geological disposal system is proposed for the high-level wastes from pyroprocessing of spent fuels. The amount and characteristics of high-level wastes are analyzed based on the material balance of pyroprocessing. Around 665 kg of monazite ceramic wastes are expected from the pyroprocessing of 10 MtU of PWR spent fuels. Decay heat from monazite ceramic wastes is calculated using the ORIGEN-ARP program. Disposal modules consisting of storage cans, overpacks, and a deposition hole or a disposal tunnel are proposed. Four kinds of deposition methods are proposed. Thermal design is carried out with ABAQUS program and geological data obtained from the KAERI Underground Research Tunnel. Through the thermal analysis, the spacing between the disposal modules

Basic study on characteristics of some important equilibrium fuel cycles of PWR

International Nuclear Information System (INIS)

Waris, A.; Sekimoto, H.

2001-01-01

Equilibrium fuel cycle characteristics of a light water reactor (LWR) with enriched uranium supply were evaluated. In this study, five kinds of fuel cycles of 3000 MWt pressurized water reactor (PWR) were investigated, and a method to determine the uranium enrichment in order to achieve their criticality was presented. The results show that the enrichment decreases considerably with increasing number of confined heavy nuclides when U is discharged from the reactor. The required natural uranium was also evaluated for two different enrichment processes. The amount of required natural uranium also decreases as well. On the other hand, when U is totally confined, the enrichment becomes unacceptably high. Furthermore, Pu and minor actinides (MA) confining seem effective to incinerate the discharged radio-toxic wastes

EdF: high tension(s) metamorphosis

International Nuclear Information System (INIS)

Roussely, F.; Arnoux, P.; Baritault, A.; Alto, P.; Castets, C.; Secondi, J.

2003-01-01

Electricite de France, the French electric utility, has to face a formidable mutation. The deregulation of the power market will lead to a social, commercial, judicial, financial and international 'big-bang'. The company has been weakened by disappointing results and by an embarrassing running into debts. This dossier analyzes the consequences of the deregulation of the French power market on the future evolution of EdF. It includes the analysis made by a French economist, E. Cohen, an interview and a portrait of F. Roussely, head of EdF, a presentation of Easenergy, a start-up of EdF which makes partnerships with US energy-related companies, the worries of EdF's employees and the redistribution of the syndicates power inside the company, the controversy around EdF's 2002 results and the points that remained in the shade, EdF's European competitors and the progressive opening of the French power market, EDF's production tool and its availability (58 nuclear reactors, 538 hydroelectric power plants and 26 thermal power plants), the costly foreign markets strategy of EdF and the under-capitalization of the company. (J.S.)

Improvements of nuclear fuel management in pressurized water reactors (PWR)

International Nuclear Information System (INIS)

Schwartz, J.P.

1978-07-01

The severe variations to which the different elements contributing to the determination of the fuel cycle cost are subjected have led to a reopening of the problem of ''optimization'' of nuclear fuel management. The increase in costs of uranium ore, isotope separation work units (swu), reprocessing, the political implications of proliferation associated with the employment of reprocessing operations have been at the origin of a reassessment of present-day management. It therefore appeared to be appropriate to study variants with respect to a reference mode represented by the management of the PWR 900 MWe systems, without burnable poison in the cycle at equilibrium (Case 3 of Table 1). In order to obtain a complete view of impacts of such modifications, computations were carried out as far as the appraisal of the cycle cost and with reprocessing. There has likewise been added to this the estimate of the gain anticipated from certain improvements in the neutron balance contributed at the level of the lattice

Seismic analysis with FEM for fuel transfer system of PWR nuclear power plant

International Nuclear Information System (INIS)

Jia Xiaofeng; Liu Pengliang; Bi Xiangjun; Ji Shunying

2012-01-01

In the PWR nuclear power plant, the function of the fuel transfer system (FTS) is to transfer the fuel assembly between the reactor building and the fuel building. The seismic analysis of the transfer system structure should be carried out to ensure the safety under OBE and SSE. Therefore, the ANASYS 12.0 software is adopted to construct the finite element analysis model for the fuel transfer system in a million kilowatt nuclear power plant. For the various configurations of FTS in the operating process, the stresses of the main structures, such as the transfer tube, fuel assembly container, fuel conveyor car, lifting frame in the reactor building, lifting frame in the fuel building, support and guide structure of conveyor car and the lifting frame in both buildings, are computed. The stresses are combined with the method of square root of square sum (SRSS) and assessed under various seismic conditions based on RCCM code, the results of the assessment satisfy the code. The results show that the stresses of the fuel transfer system structure meet the strength requirement, meanwhile, it can withstand the earthquake well. (authors)

EDF at a glance 2003

International Nuclear Information System (INIS)

2004-01-01

This document presents the organization and key data of Electricite de France (EdF) group for 2003: business, priorities, values, different branches of the group (continental Europe, western Europe, Mediterranean branch, Africa, Asia-Pacific, Americas, Dalkia energy services, development branch), management, board of directors, market opening in France, human resource strategy, R and D, customers (residential, commercial, business and industrial, local authorities), industrial tool (fossil-fired generation, network management, renewable energy sources), reliability, financial flexibility, sustainable development (role of renewables in generation, CO 2 emissions, radiation protection, nuclear waste, environment, handling fuel poverty, landscape conservation), contacts and affiliates outside France. (J.S.)

EDF, AREVA, Alstom and the others: where is the problem?

International Nuclear Information System (INIS)

2011-01-01

After having recalled the context of the Messmer plan of 1974 according to which the roles of the different actors of the nuclear sector were clearly distributed, this article describes and comments the evolution of this context, notably after the N4 threshold has been reached: the French industry has lost many experts (they are retired), the concurrence has evolved as well as relationships between clients and providers. The article outlines that a new balance between the main actors is now needed. It comments the case of the EPR, the emergence of new PWR concepts, the different roles EDF, AREVA, GDF-Suez or Total may have depending on the project (they can be client, investor, or operator). Principles for a new strategy and operation for this sector and its actors are formulated

Simulation of the fuel rod thermal hydraulic performance during the blow down phase in a PWR

International Nuclear Information System (INIS)

Gadelha, J.A.M.

1982-10-01

A digital computer code to predict the fuel rod thermalhydraulic performance during a postulated loss-of-coolant accident (LOCA) in the primary circuit of a PWR nuclear power plant is developed. The fuel rod corresponds to that in an average channel in the core. Only the blowdown phase is considered during the accident. The conservation equations of mass, momentum, and energy, and the heat conduction equation are solved to determine the fuel rod conditions during the accident. Finite differences are applied as a numerical method in the solution of the equations modelling the rod and coolant conditions. (Author) [pt

Investigation into fuel pin reshuffling options in PWR in-core fuel management for enhancement of efficient use of nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Daing, Aung Tharn, E-mail: atdaing@khu.ac.kr; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

2014-07-01

Highlights: • This paper discusses an alternative option, fuel pin reshuffling for maximization of cycle energy production. • The prediction results of isotopic compositions of each burnt pin are verified. • The operating performance is analyzed at equilibrium core with fuel pin reshuffling. • The possibility of reuse of spent fuel pins for reduction of fresh fuel assemblies is investigated. - Abstract: An alternative way to enhance efficient use of nuclear fuel is investigated through fuel pin reshuffling options within PWR fuel assembly (FA). In modeling FA with reshuffled pins, as prerequisite, the single pin calculation method is proposed to estimate the isotopic compositions of each pin of burnt FA in the core-wide environment. Subsequently, such estimation has been verified by comparing with the neutronic performance of the reference design. Two scenarios are concerned, i.e., first scenario was targeted on the improvement of the uniform flux spatial distribution and on the enhancement of neutron economy by simply reshuffling the existing fuel pins in once-burnt fuel assemblies, and second one was focused on reduction of fresh fuel loading and discharged fuel assemblies with more economic incentives by reusing some available spent fuel pins still carrying enough reactivity that are mechanically sound ascertained. In scenario-1, the operating time was merely somewhat increased for few minutes when treating eight FAs by keeping enough safety margins. The scenario-2 was proved to reduce four fresh FAs loading without largely losing any targeted parameters from the safety aspect despite loss of 14 effective full power days for operation at reference plant full rated power.

Observations of crud deposits, corrosion and erosion of BWR and PWR fuel

International Nuclear Information System (INIS)

Bairiot, H.

1983-01-01

The BWR experience is limited to one reactor but the PWR experience covers a wide range of successive generations of power plants (7 in total). The systems are described and their water chemistry briefly commented. Some R and D performed on the effects of the operating regimes (steady state and transients) are summarized. Observations made by pool-side inspections and postirradiation examinations of fuel are outlined concerning water chemistry effects (crud deposits and corrosion) and ''mechanical'' coolant-cladding interaction (chip deposits and baffle jetting). (author)

EDF Group - Annual Report 2009

International Nuclear Information System (INIS)

2010-01-01

The EDF Group is a leading player in the energy industry, active in all areas of the electricity value chain, from generation to trading and network management, with expanding operations in the natural gas chain. It has a sound business model, evenly balanced between regulated and deregulated activities. The EDF Group is the leader in the French and British electricity markets and has solid positions in Germany and Italy and numerous other European countries, as well as industrial operations in Asia and the United States. Everywhere it operates, the EDF Group is a model of quality public service for the energy sector. With fi rst-rate human resources, R and D capability and generation expertise in nuclear, fossil-fired and renewable energies, particularly hydro, together with energy eco-efficiency offers, the EDF Group delivers competitive solutions that help ensure sustainable economic development and climate protection. This document is EDF Group's annual report for the year 2009. It contains information about Group profile, governance, business, development strategy, sales and marketing, positions in Europe and international activities. The document is made of several reports: the Activity and Sustainable Development Report, the Financial Report, the Management Report, the Report by the Chairman of EDF Board of Directors on corporate governance and internal control procedures, the Milestones report, the 'EDF at a glance' report, and the Sustainable Development Indicators

EDF group - annual report 2003; Groupe EDF - rapport annuel 2003

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

This document contains the magazine, the financial statements and the sustainable development report of Electricite de France (EdF) group for 2003: 1 - the magazine (chairman's statement, group profile, vision and strategy); 2 - the consolidated financial statements for the period ended 31 December 2003 (statutory auditors' report on the consolidated financial statements, EDF's summary annual financial statements); 3 - sustainable development report (transparency and dialogue, responsibility, commitment, partnerships for progress). (J.S.)

The de-construction programme for EDF'S first generation power plants

International Nuclear Information System (INIS)

Zask, G.; Corcuff, A.

2003-01-01

Before 2001 EDF had adopted a 'long wait' scenario for the de-construction of nuclear power plants, consisting of waiting for a period of 5 to 10 years for IAEA level 2 (partial release of the site), then postponing the total de-construction of the facility for 25 to 50 years, in order to benefit from the natural decay of the radioactivity and thereby reduce the dosimetry of the personnel and the costs. Today, as regards the 8 power plants of the first generation and Creys-Malville, EDF has decided to undertake the total de-construction over a period of 25 years of all its reactors that have ceased commercial operation. The forecast cost of this programme amounts to Euro 3 billion. It thus intends to demonstrate its capacity to control, within the scale of the human lifespan, the entire life cycle of its nuclear power plants, while minimising the impact on man and the environment. This choice of immediate dismantling for the power plants of the first generation does not prejudge what will be done, when the time comes, for the PWR plants currently in operation. The paper has the following contents: 1. General; 2.The facilities concerned; 3. Chronological sequence adopted for de-construction; 4. Milestones and critical paths; 4.1. Simplification of the administrative procedures; 4.2. The waste channels; 4.3. Industrial organisation; 4.3.1. Creation of Centre d'Ingenierie De-construction et Environnement (CIDEN); 4.3.2. Responsibilities, interfaces and co-ordination of the various departments concerned (Role of owner: Responsibility as nuclear operator; Overall co-ordination; Operational steering of projects); 4.3.3. Internal authorization process; 4.3.4. Obtaining and maintaining skills (Training; Environment; De-construction; Radiological protection, radiology, radioecology; Processing of waste and approval of the cycle); 4.3.5. Monitoring of works and surveillance of service providers; 4.3.6. Radiological protection, environment; 4.3.7. Control of conventional

Research of natural resources saving by design studies of Pressurized Light Water Reactors and High Conversion PWR cores with mixed oxide fuels composed of thorium/uranium/plutonium

International Nuclear Information System (INIS)

Vallet, V.

2012-01-01

Within the framework of innovative neutronic conception of Pressurized Light Water Reactors (PWR) of 3. generation, saving of natural resources is of paramount importance for sustainable nuclear energy production. This study consists in the one hand to design high Conversion Reactors exploiting mixed oxide fuels composed of thorium/uranium/plutonium, and in the other hand, to elaborate multi-recycling strategies of both plutonium and 233 U, in order to maximize natural resources economy. This study has two main objectives: first the design of High Conversion PWR (HCPWR) with mixed oxide fuels composed of thorium/uranium/plutonium, and secondly the setting up of multi-recycling strategies of both plutonium and 233 U, to better natural resources economy. The approach took place in four stages. Two ways of introducing thorium into PWR have been identified: the first is with low moderator to fuel volume ratios (MR) and ThPuO 2 fuel, and the second is with standard or high MR and ThUO 2 fuel. The first way led to the design of under-moderated HCPWR following the criteria of high 233 U production and low plutonium consumption. This second step came up with two specific concepts, from which multi-recycling strategies have been elaborated. The exclusive production and recycling of 233 U inside HCPWR limits the annual economy of natural uranium to approximately 30%. It was brought to light that the strong need in plutonium in the HCPWR dedicated to 233 U production is the limiting factor. That is why it was eventually proposed to study how the production of 233 U within PWR (with standard MR), from 2020. It was shown that the anticipated production of 233 U in dedicated PWR relaxes the constraint on plutonium inventories and favours the transition toward a symbiotic reactor fleet composed of both PWR and HCPWR loaded with thorium fuel. This strategy is more adapted and leads to an annual economy of natural uranium of about 65%. (author) [fr

Comparison of thermal behavior of different PWR fuel rod simulators for LOCA experiments

International Nuclear Information System (INIS)

Casal, V.; Malang, S.; Rust, K.

1982-10-01

For experimental investigations of a loss-of-coolant accident (LOCA) of a PWR electrical heater rods are applied as thermal fuel rod simulators. To substitute heater rods from the SEMISCALE program by INTERATOM-KfK heater rods in a current experimental program at the Instituut for Energiteknikk-(OECD-Halden), the thermodynamic behavior of different heater rods during a LOCA were compared. The results show, that SEMISCALE-heater rods can be replaced by those fabricated by INTERATOM. (orig.) [de

Small PWR 'PFPWR50' using cermet fuel of Th-Pu particles

International Nuclear Information System (INIS)

Hirayama, Takashi; Shimazu, Yoichiro

2009-01-01

An innovative concept of PFPWR50 has been studied. The main feature of PFPWR50 has been to adopt TRISO coated fuel particles in a conventional PWR cladding. Coated fuel particle provides good confining ability of fission products. But it is pointed out that swelling of SiC layer at low temperature by irradiation has possibilities of degrading the integrity of coated fuel particle in the LWR environment. Thus, we examined the use of Cermet fuel replacing SiC layer to Zr metal or Zr compound. And the nuclear fuel has been used as fuel compact, which is configured to fix coated fuel particles in the matrix material to the shape of fuel pellet. In the previous study, graphite matrix is adopted as the matrix material. According to the burnup calculations of the several fuel concepts with those covering layers, we decide to use Zr layer embedded in Zr metal base or ZrC layer with graphite matrix. But carbon has the problem at low temperature by irradiation as well as SiC. Therefore, Zr covering layer and Zr metal base are finally selected. The other feature of PFPWR50 concept has been that the excess reactivity is suppressed during a cycle by initially loading burnable poison (gadolinia) in the fuels. In this study, a new loading pattern is determined by combining 7 types of assemblies in which the gadolinia concentration and the number of the fuel rods with gadolinia are different. This new core gives 6.7 equivalent full power years (EFPY) as the core life of a cycle. And the excess reactivity is suppressed to less than 2.0%Δk/k during the cycle. (author)

Measurement of the gap and grain boundary inventories of Cs, Sr and I in domestic used PWR fuels

International Nuclear Information System (INIS)

Kim, S. S.; Choi, J. W.; Seo, H. S.; Cho, W. J.; Kang, K. C.; Kwon, S. H.

2007-01-01

Inventories of soluble elements in the gap and grain boundaries of domestic used PWR fuel pellets were measured to estimate the quantities of radionuclides that are liable to be rapidly released into the groundwater of a disposal site. The gap inventory of cesium for the pellets in the used fuel with a burn-up range of 45 to 66 GWD/MTU showed 0.85 to 1.7% of its total inventory, which was close to 1/6 to 1/3 of the fission gas release fraction (FGRF). However, the amounts of cesium released from the gaps of the pellets below 40 GWD/MTU of a burn-up and less than 1% FGRF were so erratic that the gap inventory could not be defined by its FGRF. Strontium inventories in the gap and grain boundaries of the pellets in the same rod were not significantly varied, and the iodine inventory in the gap of the used PWR fuels was estimated to be less than or the same as the FGRF

Ageing management of EDF NPP concrete structures: from the theory to the reality

International Nuclear Information System (INIS)

Gallitre, E.; Brunet, C.

2009-01-01

This paper presents EDF practice about concrete structure ageing management, from the materials analysis to the formal procedure which allows EDF to increase 900 MWe NPP lifetime until 40 years. This practice is based on a methodology which identifies every ageing mechanism, then both plants feedback and state of the art are screened and conclusions are drawn up and taken into account into an 'ageing analysis data sheet'. That leads to a collection of 57 data sheets: if the data sheet conclusion is not clear enough, then a more detailed report is launched. This ageing management detailed report gathers every theoretical knowledge and monitoring data; its objective is to propose a solution for ageing management: this solution can include more inspections or specific research development. After a first practise on the 34 French 900 MWe PWR units, only 2 generic ageing management detailed reports have been needed for the civil engineering part: one about reactor building containment which focuses on two mechanisms: steel liner corrosion and pre-stress losses, and one about other structures which focuses on Alkali-Aggregate Reaction (AAR). (authors)

Development of a Computer Program for an Analysis of the Logistics and Transportation Costs of the PWR Spent Fuels in Korea

International Nuclear Information System (INIS)

Cha, Jeong Hun; Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won

2009-01-01

It is expected that a substantial amount of spent fuels will be transported from the four nuclear power plant (NPP) sites in Korea to a hypothetical centralized interim storage facility or a final repository in the near future. The cost for the transportation is proportional to the amount of spent fuels. In this paper, a cost estimation program is developed based on the conceptual design of a transportation system and a logistics analysis. Using the developed computer program, named as CASK, the minimum capacity of a centralized interim storage facility (CISF) and the transportation cost for PWR spent fuels are calculated. The PWR spent fuels are transported from 4 NPP sites to a final repository (FR) via the CISF. Since NPP sites and the CISF are located along the coast, a sea-transportation is considered and a road-transportation is considered between the CISF and the FR. The result shows that the minimum capacity of the interim storage facility is 15,000 MTU

Evaluation of the heat transfer in a geological repository concept containing PWR, VHTR and hybrid ads-fission spent fuels

Energy Technology Data Exchange (ETDEWEB)

Jonusan, Raoni A.S.; Pereira, Fernando; Velasquez, Carlos E.; Salome, Jean A.D.; Cardoso, Fabiano; Pereira, Claubia; Fortini, Angela, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-11-01

The investigation of the thermal behavior of spent fuel (SF) materials is essential to determining appropriate potential sites to accommodate geological repositories as well as the design of canisters, considering their potential risk to people health and of environmental contamination. This work presents studies of the temperature in a canister containing spent fuels discharged from Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) reactor systems in a geological repository concept. The thermal analyses were performed with the software ANSYS, which is widely used to solve engineering problems through the Finite Element Method. The ANSYS Transient Thermal module was used. The spent nuclear fuels were set as heat sources using data of previous studies derived from decay heat curves. The studies were based on comparison of the mean temperature on a canister surface along the time under geological disposal conditions, for a same amount of each type of spent nuclear fuel evaluated. The results conclude that fuels from VHTR and ADS systems are inappropriate to be disposed in a standardized PWR canister, demanding new studies to determine the optimal amount of spent fuel and new internal canister geometries. It is also possible to conclude that the hypothetical situation of a single type of canister being used to accommodate different types of spent nuclear fuels is not technically feasible. (author)

Fuel rod-to-support contact pressure and stress measurement for CHASNUPP-1(PWR) fuel

International Nuclear Information System (INIS)

Waseem; Elahi, N.; Siddiqui, A.; Murtaza, G.

2011-01-01

Research highlights: → A detailed finite element model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. → The spring hold-down force is calculated using the contact pressure obtained from the FE model. → Experiment has also been conducted in the same environment for the measurement of this force. → The spring hold-down force values obtained from both studies confirm the validation of this analysis. → The stress obtained through this analysis is less than the yield strength of spacer grid material, thus fulfils the structural integrity criteria of grid. - Abstract: This analysis has been made in an attempt to measure the contact pressure of the PWR fuel assembly spacer grid spring and to verify its structural integrity at room temperature in air. A detailed finite element (FE) model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The FE model of a fuel rod-to-support system is produced with shell and contact elements. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies are compared, which show good agreement, and in turn confirm the validation of this analysis. The Stress obtained through this analysis is less than the yield strength of spacer grid material (Inconel-718), thus fulfils the structural integrity criteria of grid.

Fuel rod-to-support contact pressure and stress measurement for CHASNUPP-1(PWR) fuel

Energy Technology Data Exchange (ETDEWEB)

Waseem, E-mail: wazim_me@hotmail.co [Directorate General Nuclear Power Fuel, Pakistan Atomic Energy Commission, P.O. Box No. 1847, Islamabad 44000 (Pakistan); Elahi, N.; Siddiqui, A.; Murtaza, G. [Directorate General Nuclear Power Fuel, Pakistan Atomic Energy Commission, P.O. Box No. 1847, Islamabad 44000 (Pakistan)

2011-01-15

Research highlights: A detailed finite element model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies confirm the validation of this analysis. The stress obtained through this analysis is less than the yield strength of spacer grid material, thus fulfils the structural integrity criteria of grid. - Abstract: This analysis has been made in an attempt to measure the contact pressure of the PWR fuel assembly spacer grid spring and to verify its structural integrity at room temperature in air. A detailed finite element (FE) model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The FE model of a fuel rod-to-support system is produced with shell and contact elements. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies are compared, which show good agreement, and in turn confirm the validation of this analysis. The Stress obtained through this analysis is less than the yield strength of spacer grid material (Inconel-718), thus fulfils the structural integrity criteria of grid.

Open and closed fuel cycle of HWR and PWR. How large is the high-level radioactive wastes repository; Ciclos de combustible abierto y cerrado con HWR y PWR. ?Cuanto mas grande es el repositorio de residuos radiactivos de alta actividad?

Energy Technology Data Exchange (ETDEWEB)

Bevilacqua, Arturo M. [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche

1996-07-01

A conceptual analysis was carried out on the size of a high-level wastes (HLW) repository for the waste arising from once-through and closed fuel cycles with (HLW) and PWR. The mass, the activity and thermal loading was calculated with the ORIGEN2.1 computer code for the spent fuel and for the high-level liquid wastes. It was considered a minimum burnup of 7.000 MW.d/t U and 33.000 MW.d/t U for HWR and PWR respectively, cooling times of 20 and 55 years, reprocessing recovery ratios of 99% and 99.7% and a total electricity production of 81.6 GW(e).a. It was concluded that the cooling time is the most important repository size reproduction parameter for the closed cycles. On the other hand, the spent fuel mass for the once-through cycles does not depend on the cooling time what prevents repository size reduction once a cooling time of 55 years is reached. The repository size reduction in the case of HWR is larger than in the case of PWR, owing to the larger fuel mass required to produce the specific electricity amount. (author)

EDF group - annual report 2003; Groupe EDF - rapport annuel 2003

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

This document contains the magazine, the financial statements and the sustainable development report of Electricite de France (EdF) group for 2003: 1 - the magazine (chairman's statement, group profile, vision and strategy); 2 - the consolidated financial statements for the period ended 31 December 2003 (statutory auditors' report on the consolidated financial statements, EDF's summary annual financial statements); 3 - sustainable development report (transparency and dialogue, responsibility, commitment, partnerships for progress). (J.S.)

The Effect of Material Homogenization in Calculating the Gamma-Ray dose from Spent PWR Fuel Pins in an Air Medium

International Nuclear Information System (INIS)

TH Trumbull

2005-01-01

The effect of material homogenization on the calculated dose rate was studied for several arrangements of typical PWR spent fuel pins in an air medium using the Monte Carlo code, MCNP. The models analyzed increased in geometric complexity, beginning with a single fuel pin, progressing to ''small'' lattices, i.e., 3x3, 5x5, 7x7 fuel pins, and culminating with a full 17x17 pin PWR bundle analysis. The fuel pin dimensions and compositions were taken directly from a previous study and efforts were made to parallel this study by specifying identical flux-to-dose functions and gamma-ray source spectra. The analysis shows two competing components to the overall effect of material homogenization on calculated dose rate. Homogenization of pin lattices tends to lower the effect of radiation ''channeling'' but increase the effect of ''source redistribution.'' Depending on the size of the lattice and location of the detectors, the net effect of material homogenization on dose rate can be insignificant or range from a 6% decrease to a 35% increase relative to the detailed geometry model

Numerical solution of the elastic non-axial contact between pellet and cladding of fuel rod in PWR

International Nuclear Information System (INIS)

Zymak, J.

1987-08-01

Elastic non-axial contacts between the pellet and the cladding of a fuel rod in a pressurized water reactor were calculated. The existence and the uniqueness of the solution were proved. The problem was approximated by the finite element method and quadratic programming was used for the solution. The results will be used in the solution of the probabilistic model of a fuel rod with non-axial pellets in a PWR. (author). 10 figs., 4 tabs., 10 refs

Status of developing advanced PWR in Japan

International Nuclear Information System (INIS)

Iida, Yotaro

1982-01-01

During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

Comparison of problems and experience of core operation with distorted fuel element assemblies in VVER-1000 and PWR reactors

International Nuclear Information System (INIS)

Afanas'ev, A.

1999-01-01

The main reactors leading to distortion of fuel element assemblies during reactor operation were studied. A series of actions which compensate this effect was proposed. Criteria of operation limitation in VVER-1000 and PWR reactors are described

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

Parallel GPU implementation of PWR reactor burnup

International Nuclear Information System (INIS)

Heimlich, A.; Silva, F.C.; Martinez, A.S.

2016-01-01

Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

A comparison of Zircaloy oxide thicknesses on Millstone-3 and North Anna-1 PWR fuel cladding

International Nuclear Information System (INIS)

Polley, M.V.; Evans, H.E.

1993-08-01

High concentrations of lithium in the coolant may enhance the corrosion rate of Zircaloy fuel cladding. In the present work, oxide thicknesses on fuel cladding from the Millstone 3 PWR were compared with those from the North Anna 1 PWR. The intention was to identify whether the higher lithium levels (up to 3.5 ppM) in the Millstone 3 primary coolant during cycles 2 and 3 led to significantly greater oxidation rates than in North Anna 1 which operated generally with lithium levels lower than 2.2 ppM. The comparisons were made by comparing the measurements with code predictions of Zircaloy oxidation in order to factor out the effect of operational variables on the oxide thicknesses achieved. Overall, Millstone 3 oxide thicknesses were found to be approximately 14% greater than North Anna 1 values. However, approximately 29% lower oxide thicknesses were found on reload Millstone 3 rods exposed to one cycle of elevated lithium chemistry than on Millstone 3 initial fuel exposed to one cycle of normal lithium chemistry during cycle 1. Furthermore, oxide thicknesses on Millstone 3 rods exposed to two cycles of elevated lithium chemistry were approximately 36% lower than on Millstone 3 rods exposed to one cycle of normal lithium chemistry plus one cycle of elevated lithium chemistry. Therefore, it cannot be concluded that elevated lithium operation in Millstone 3 led to enhanced Zircaloy fuel clad corrosion

Characterization and modeling of the thermal hydraulic and chemical environment of fuel claddings of PWR reactors during boiling

International Nuclear Information System (INIS)

March, Ph.

1999-01-01

In pressurised water reactors (PWR), nucleate boiling can strongly influence the oxidation rate of the fuel cladding. To improve our understanding of the effect of the boiling phenomenon on corrosion kinetics, information about the chemical and thermal hydraulic boundary conditions at the heating rod surface is needed. Moreover, very few data are available in the range of thermal hydraulic parameters of PWR cores (15,5 MPa and 340 deg C) concerning the two-phase flow pattern close to the fuel cladding. A visualization device has been adapted on an out-of-pile loop Reggae to obtain both qualitative and quantitative data. These observations provide a direct access to the geometrical properties of the vapor inclusions, the onset of nucleate boiling and the gas velocity and trajectory. An image processing method has been validated to measure both void fraction and interfacial area concentration in a bubbly two-phase flow. Thus, the visualization device proves to be a suitable and accurate instrumentation to characterize nucleate boiling in PWR conditions. The experimental results analysis indicates that a local approach is needed for the modelling of the fuel rod chemical environment. To simulate the chemical additives enrichment, a new model is proposed where the vapor bubbles are now considered as physical obstacles for the liquid access to the rod surface. The influence of the two-phase flow pattern appears to be of major importance for the enrichment phenomenon. This study clearly demonstrates the existence of strong interactions between the two-phase flow pattern, the rod surface condition, the corrosion process and the water chemistry. (author)

Prototypical fabrication of PWR spent fuel shipping cask

International Nuclear Information System (INIS)

Kwack, Eun Ho; Kim, Byung Ku; Kang, Hee Yung; Lee, Chung Young; Jeon, Kyeong Lak; Lee, Bum Soo

1985-02-01

This report describes about the safety analysis for the spent fuel shipping cask, which is used to transfer a single fuel assembly discharged from PWR in operation in Korea. The contents cover the methods and the results of structural, thermal, thermo-hydraulic, radiation shield and criticality detail analysis. The safety evaluation has been made under the normal transportation and hypothetical accident conditions such as 30ft free drop, puncture, fire, immersion, penetration, corner drop, etc,. Some corrections in design are made, and a brief information for fabrication and transportation are obtained by the use of a 1/6 scale model. The design is based on one year cooling time of the spent fuel with 40,000 MWT/MTU maximum burnup, which gives 7.2KW decay heat and 1.6x10 6 ci/hr radiation intensity. The cask is composed of main body with the double closures, impact limiter and fuel basket. The inner shell, inner closure and valves constitute the pressure boundary of the containment. The inner, intermediate and outer shells, upper and lower forgings are made of stainless steel which compose the main body with lead for gamma shield and 50% ethylene glycol for neutron shield. The impact limiters are made of balsa wood on both end sides of the cask to protect the cask from a sudden shocks in accident during the transportation. The analysis results show that the cask is proved to retain its structural integrity within allowable stress and to be safe under the normal and hypothetical accident conditions, and the maximum dose rates of radiation at 2m distance from the surface of the cask are less than the required values. The weight will be 23.2tons in dry and 27.8 tons in wet with fuel loaded. All the design data, calculated results for the structural integrity, shield and thermal analysis are shown in this report with the basic drawings. (Author)

HEXBU-3D, a three-dimensional PWR-simulator program for hexagonal fuel assemblies

International Nuclear Information System (INIS)

Karvinen, E.

1981-06-01

HEXBU-3D is a three-dimensional nodal simulator program for PWR reactors. It is designed for a reactor core that consists of hexagonal fuel assemblies and of big follower-type control assemblies. The program solves two-group diffusion equations in homogenized fuel assembly geometry by a sophisticated nodal method. The treatment of feedback effects from xenon-poisoning, fuel temperature, moderator temperature and density and soluble boron concentration are included in the program. The nodal equations are solved by a fast two-level iteration technique and the eigenvalue can be either the effective multiplication factor or the boron concentration of the moderator. Burnup calculations are performed by tabulated sets of burnup-dependent cross sections evaluated by a cell burnup program. HEXBY-3D has been originally programmed in FORTRAN V for the UNIVAC 1108 computer, but there is also another version which is operable on the CDC CYBER 170 computer. (author)

A probabilistic analysis of PWR and BWR fuel rod performance using the code CASINO-SLEUTH

International Nuclear Information System (INIS)

Bull, A.J.

1987-01-01

This paper presents a brief description of the Monte Carlo and response surface techniques used in the code, and a probabilistic analysis of fuel rod performance in PWR and BWR applications. The analysis shows that fission gas release predictions are very sensitive to changes in certain of the code's inputs, identifies the most dominant input parameters and compares their effects in the two cases. (orig./HP)

Mixed PWR core loadings with inert matrix Pu-fuel assemblies

International Nuclear Information System (INIS)

Stanculescu, A.; Kasemeyer, U.; Paratte, J.-M.; Chawla, R.

1999-01-01

The most efficient way to enhance plutonium consumption in light water reactors is to eliminate the production of plutonium all together. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. At PSI, studies have focused on employing ZrO 2 as inert matrix. Adding a burnable poison to such a fuel proves to be necessary. As a result of scoping studies, Er 2 O 3 was identified as the most suitable burnable poison material. The results of whole-core three-dimensional neutronics analyses indicated, for a present-day 1000 MW e pressurised water reactor, the feasibility of an asymptotic equilibrium four-batch cycle fuelled solely with the proposed PuO 2 -Er 2 O 3 -ZrO 2 inert matrix fuel (IMF). The present paper presents the results of more recent investigations related to 'real-life' situations, which call for transition configurations in which mixed IMF and UO 2 assembly loadings must be considered. To determine the influence of the introduction of IMF assemblies on the characteristics of a UO 2 -fuelled core, three-dimensional full-core calculations have been performed for a present-day 1000 MW e PWR containing up to 12 optimised IMF assemblies. (author)

Fuel management optimization for a PWR

International Nuclear Information System (INIS)

Dumas, M.; Robeau, D.

1981-04-01

This study is aimed to optimize the refueling pattern of a PWR. Two methods are developed, they are based on a linearized form of the optimization problem. The first method determines a feasible solution in two steps; in the first one the original problem is replaced by a relaxed one which is solved by the Method of Approximation Programming. The second step is based on the Branch and Bound method to find the feasible solution closest to the solution obtained in the first step. The second method starts from a given refueling pattern and tries to improve this pattern by the calculation of the effects of 2 by 2, 3 by 3 and 4 by 4 permutations on the objective function. Numerical results are given for a typical PWR refueling using the two methods

A calculation methodology applied for fuel management in PWR type reactors using first order perturbation theory

International Nuclear Information System (INIS)

Rossini, M.R.

1992-01-01

An attempt has been made to obtain a strategy coherent with the available instruments and that could be implemented with future developments. A calculation methodology was developed for fuel reload in PWR reactors, which evolves cell calculation with the HAMMER-TECHNION code and neutronics calculation with the CITATION code.The management strategy adopted consists of fuel element position changing at the beginning of each reactor cycle in order to decrease the radial peak factor. The bi-dimensional, two group First Order perturbation theory was used for the mathematical modeling. (L.C.J.A.)

Castor-V/21 PWR spent fuel storage cask performance test

International Nuclear Information System (INIS)

Creer, J.M.; Schoonen, D.H.

1986-01-01

Performance testing of a CASTOR-V/21 PWR spent fuel storage cask manufactured by Gesellschaft fur Nuklear Service (GNS) was performed as part of a cooperative program between Virginia Power and the US Department of Energy. The performance test consisted of obtaining cask handling experience and heat transfer, shielding, and limited fuel integrity data. Five heat transfer test runs were performed with 21 Surry reactor spent fuel assemblies generating approximately 28 kW. Test runs were performed vacuum, nitrogen, and helium backfill environments with the cask in both vertical and horizontal orientations. Cask exterior surface gamma and neutron dose rates were measured with the cask fully loaded. Gas samples were obtained at the beginning and end of each run with nitrogen or helium environments to verify fuel integrity. The heat transfer performance of the CASTOR-V/21 cask was exceptionally good. Peak clad temperatures with helium and nitrogen environments with the cask in a vertical orientation and with helium with the cask in a horizontal orientation were less than 380 0 C. Vertical vacuum and horizontal nitrogen test runs resulted in peak clad temperatures over 380 0 , but the temperatures were not excessively high ( 0 C). The shielding performance of the cask met the design goal of less than 200 mrem/hr. Cask surface dose rates of <75 mrem/hr can easily be established with minor gamma shielding design refinements if desired. Gas samples obtained during testing indicated no leaking fuel rods were present in the cask. It was concluded that the cask performed satisfactorily from heat transfer and shielding perspectives

CASTOR-V/21 PWR spent fuel storage cask performance test

International Nuclear Information System (INIS)

Creer, J.M.; Schoonen, D.H.

1986-01-01

Performance testing of a CASTOR-V/21 PWR spent fuel storage cask manufactured by Gesellschaft fur Nuklear Service (GNS) was performed as part of a cooperative program between Virginia Power and the US Department of Energy. The performance test consisted of obtaining cask handling experience and heat transfer, shielding, and limited fuel integrity data. Five heat transfer test runs were performed with 21 Surry reactor spent fuel assemblies generating approximately 28 kW. Test runs were performed with vacuum, nitrogen, and helium backfills in both vertical and horizontal orientations. Cask exterior surface gamma and neutron dose rates were measured with the cask fully loaded. Gas samples were obtained at the beginning and end of each run with nitrogen or helium backfills to verify fuel integrity. The heat transfer performance of the CASTOR-V/21 cask was exceptionally good. Peak clad temperatures with helium and nitrogen backfills in a vertical orientation and with helium in a horizontal orientation were less than 380 0 C. Vertical vacuum and horizontal nitrogen runs resulted in peak clad temperatures over 380 0 , but the temperatures were not excessively high ( 0 C). The shielding performance of the cask met the design expectation of less than 200 mrem/h. Cask surface dose rates of <75 mrem/h can easily be established with minor gamma shielding design refinements if desired. Gas samples obtained during testing indicated no leaking fuel rods were present in the cask. It was concluded that the cask performed satisfactorily from heat transfer and shielding perspectives

PWR burnable absorber evaluation

International Nuclear Information System (INIS)

Cacciapouti, R.J.; Weader, R.J.; Malone, J.P.

1995-01-01

The purpose of the study was to evaluate the relative neurotic efficiency and fuel cycle cost benefits of PWR burnable absorbers. Establishment of reference low-leakage equilibrium in-core fuel management plans for 12-, 18- and 24-month cycles. Review of the fuel management impact of the integral fuel burnable absorber (IFBA), erbium and gadolinium. Calculation of the U 3 O 8 , UF 6 , SWU, fuel fabrication, and burnable absorber requirements for the defined fuel management plans. Estimation of fuel cycle costs of each fuel management plan at spot market and long-term market fuel prices. Estimation of the comparative savings of the different burnable absorbers in dollar equivalent per kgU of fabricated fuel. (author)

Feedback from dismantling operations (level 2) on EDF's first generation reactors

International Nuclear Information System (INIS)

West, J P.; Dionisio-Gomes, A.; Kus, J P.; Mervaux, P.; Bernet, P.; Dalmas, R.

2003-01-01

EDF's policy as regards the dismantling of the reactors that have ceased commercial operation, namely the eight power plants of the first generation and the Creys-Malville power plant, is explained. Generally speaking, prior to the year 2001, EDF had opted for the de-construction of these power plants to comply with a 'long wait' scenario, which consisted of waiting for a period of 5 to 10 years to achieve IAEA level 2 (partial release of the site), then postponing the total de-construction of the facilities for 25 to 50 years. Today, EDF has decided to undertake the total de-construction of these reactors, which have ceased commercial operation, over a period of 25 years. The purpose of this document is to present: - The reactors concerned, their background and their 'regulatory' situation, - The main operations performed and/or currently in progress, - The main elements of feedback from such operations, shedding light on the approach adopted in 2001. The installations concerned by the de-construction programme are as follows: - The 8 power plants of the first generation, which were built during the fifties and sixties and ceased commercial operation between 1973 and 1994, namely: Brennilis (industrial prototype using heavy water technology, jointly operated by EDF and CEA), the 6 power units of the NUGG type (natural uranium gas graphite) at Chinon, Saint-Laurent des Eaux and Bugey and the PWR reactor at Chooz A, - The storage silos at Saint-Laurent, where the sleeves for the fuel assemblies of reactors SLA1 and SLA2 are stored, corresponding to approximately 2000 tonnes of graphite, - The Creys-Malville reactor, FBR (fast breeder reactor) shut down in accordance with a government decision, which is currently undergoing decommissioning. At the current stage, our feedback from the dismantling operations carried out on nuclear facilities is based on (i) the work carried out or in progress that will make it possible to achieve the equivalent of IAEA level 2 in the

Conceptual design of simplified PWR

International Nuclear Information System (INIS)

Tabata, Hiroaki

1996-01-01

The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Having no intention of constructing medium-sized plants as a next generation standard plant, Japanese utilities are interested in applying passive technologies to large ones. So, Japanese utilities have studied large passive plants based on AP600 and SBWR as alternative future LWRs. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, JAPC and Japanese Utilities started the study to modify AP600 and SBWR, in order to accommodate the Japanese requirements. During a six year program up to 1994, basic concepts for 1000 MWe class Simplified PWR (SPWR) and Simplified BWR (SBWR) were developed, though there still remain several areas to be improved. These studies have now stepped into the phase of reducing construction cost and searching for maximum power rating that can be attained by reasonably practical technology. These results also suggest that it is hopeful to develop a large 3-loop passive plant (∼1200 MWe). Since Korea mainly deals with PWR, this paper summarizes SPWR study. The SPWR is jointly studied by JAPC, Japanese PWR Utilities, EdF, WH and Mitsubishi Heavy Industry. Using the AP-600 reference design as a basis, we enlarged the plant size to 3-loops and added engineering features to conform with Japanese practice and Utilities' preference. The SPWR program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1000 MWe and 3 loops. (J.P.N.)

Radionuclide release from PWR fuels in a reference tuff repository groundwater

International Nuclear Information System (INIS)

Wilson, C.N.; Oversby, V.M.

1985-03-01

The Nevada Nuclear Waste Storage Investigations Project (NNWSI) is studying the suitability of the welded devitrified Topopah Spring tuff at Yucca Mountain, Nye County, Nevada, for potential use as a high-level nuclear waste repository. In support of the Waste Package task of NNWSI, tests have been conducted under ambient air environment to measure radionuclide release from two pressurized water reactor (PWR) spent fuels in water obtained from the J-13 well near the Yucca Mountain site. Four specimen types, representing a range of fuel physical conditions that may exist in a failed waste canister containing a limited amount of water were tested. The specimen types were: fuel rod sections split open to expose bare fuel particles; rod sections with water-tight end fittings with a 2.5-cm long by 150-μm wide slit through the cladding; rod sections with water-tight end fittings and two 200-μm-diameter holes through the cladding; and undefected rod segments with water-tight end fittings. Radionuclide release results from the first 223-day test runs on H.B. Robinson spent fuel specimens in J-13 water are reported and compared to results from a previous test series in which similar Turkey Point reactor spent fuel specimens were tested on deionized water. Selected initial results are also given for Turkey Point fuel specimens tested on J-13 water. Results suggest that the actinides Pu, Am, Cm and Np are released congruently with U as the UO 2 spent fuel matrix dissolves. Fractional release of 137 Cs and 99 Tc was greater than that measured for the actinides. Generally, lower radionuclide releases were measured for the H.B. Robinson fuel in J-13 water than for Turkey Point Fuel in deionized water. 8 references, 7 figures, 9 tables

Development of computational methods to describe the mechanical behavior of PWR fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Wanninger, Andreas; Seidl, Marcus; Macian-Juan, Rafael [Technische Univ. Muenchen, Garching (Germany). Dept. of Nuclear Engineering

2016-10-15

To investigate the static mechanical response of PWR fuel assemblies (FAs) in the reactor core, a structural FA model is being developed using the FEM code ANSYS Mechanical. To assess the capabilities of the model, lateral deflection tests are performed for a reference FA. For this purpose we distinguish between two environments, in-laboratory and in-reactor for different burn-ups. The results are in qualitative agreement with experimental tests and show the stiffness decrease of the FAs during irradiation in the reactor core.

Results of a recent crud/corrosion fuel risk assessment at a U.S. PWR

International Nuclear Information System (INIS)

Lamanna, Larry; Pop, Mike; Gregorich, Carola; Harne, Richard; Jones, John

2012-09-01

In order to avoid potential fuel reliability issues, specifically crud-related issues, it is necessary to achieve and maintain a crud safe environment. Therefore, the ability to confidently predict risks associated with crud deposition on fuel becomes critically important. AREVA is applying its cutting-edge PWR Fuel Crud (Primary System corrosion products)/Corrosion Tools, i.e. COBRA-FLX (subchannel-by-subchannel T/H tool) coupled with FDIC (crud deposition tool) to subsequently perform PWR Fuel Crud /Corrosion risk assessments for operating plants in the US. After describing the method, the result of one of these assessments is presented for an operating plant in the US that has experienced recent crud observations/concerns. Both Crud Induced Localized Corrosion (CILC) and Crud Induced Power Shift (CIPS) risk assessment methods, as applied to the upcoming cycle (Cycle N), were compared to the current/on-going cycle (Cycle N-1) and to the previous cycle (Cycle N-2). The results allowed the Utility to consider crud risk management changes associated with the upcoming cycle (Cycle-N). Benchmarking of the AREVA tools, using the plant-specific crud information gained from the crud sampling/characterization for the Unit will be presented. The CIPS analysis references boron loading and the amount of insoluble iron-nickel-borates predicted for Cycles N-2, N-1, and N. The results of the CILC evaluation reference FDIC-predicted crud thickness, cladding temperature under deposit, evolution of CILC bearing species and lithium concentration in the zirconium oxide layer. The approach taken by AREVA during the evaluation was to consider both 'risk' and 'margin' to fuel performance impact caused by crud deposits. The conclusion of the assessment, illustrated by the results presented in this paper, is that the example Plant has sufficient margin in worst case conditions for CIPS and CILC risk in Cycle N, based on Cycle N-1 and Cycle N-2 conditions and behavior

EDF, the 2003-2007 project

International Nuclear Information System (INIS)

2005-01-01

This brochure presents the 2003-2007 project of Electricite de France (EdF) in the framework of the opening of European energy markets. The aim of this project is to take over the present day contract of EdF with the French government in order to become an anonymous company. The wish of F. Roussely, president of EdF, is to open and increase the capital of the company but also to fulfill its public utility missions. The stake of EdF is to stay competitive, to become a European energy leader, to stay an integrated company combining production, transport, distribution and commercialization, and finally to maintain its historical public utility standards: specific tariff conditions and services for low income people, financing of renewable energy projects. (J.S.)

Method and Result of Experiment for Support of Technical Solutions in the Field of Perfection of a Nuclear Fuel Cycle for Future PWR Reactors

International Nuclear Information System (INIS)

Ostrovskiy, V.; Kudryavtsev, E.; Tutnov, I.

2011-01-01

The paper presents the basics of approach of planning and carrying out of experiments to validate safety PWR reactors of the future when accepting technical solutions concerning using of improved fuel rods in fuel assembly. Basic principles and criteria used for the validation of technical solutions and developments in improving of nuclear fuel cycle of PWR reactors of the future are presented from the point of safety of future operation of modified fuel rods. We explore the questions of safety operation of PWR reactors with fuel assemblies, containing fuel rods with different length of fuel. The paper discusses the ways of solving of important tasks of critical facility experiments conducting for verification of new technical solutions in the sphere of PWR nuclear fuel cycle improvement on the base of international standards ISO 2000:9000 and functional safety recommendations of IEC (International Electromechanical Commission). New Federal laws of Russian Federation define the main principle for demands to NPP and any supplier of nuclear techniques. The principle is 'quantity indicators of risk should not exceed comprehensible social size of the established indicators of safety for any moment of operation of NPP'. On the other hand the second principle should be applied to extraction of the greatest benefit from operation of the equipment, systems or the NPP as whole: 'The long operation and full commercial use of resource and service properties of the equipment, systems and the NPP as a whole'. Realization of this principle assumes development and introduction of new technical solutions for a validation of guarantees of safety of the future operation of NPP or it separate components. Solving the practical problems of a validation of safety use of fuel rods with the increased length of a fuel column in fuel assembly in nuclear reactors of the future, we should choose new strategies and programs of verification experiments on the base of the analysis of guarantees

2003 annual results of EdF group

International Nuclear Information System (INIS)

2004-03-01

The Electricite de France (EdF) group Board of Directors, meeting on March 11, 2004, under the Chairmanship of Francois Roussely, reviewed the audited consolidated financial statements for the year ended 12/31/2003. This document presents the consolidated results of EdF group for 2003: consolidated financial statements, highlights of the year, focus on 2003 events, commercial results, EDF in Europe and worldwide, EDF France highlights, key figures. (J.S.)

Development of automated generation system of accidental operating procedures for a PWR

International Nuclear Information System (INIS)

Artaud, J.L.

1991-06-01

The aim of the ACACIA project is to develop an automated generation system of accident operating procedures for a PWR. This research and development study, common at CEA and EDF, has two objectives: at mean-dated the realization of a validation tool and a procedure generation; at long-dated the dynamic generation of real time procedures. This work is consecrated at the realization of 2 prototypes. These prototypes and the technics used are described in detail. The last chapter explores the perspectives given by this type of tool [fr

The KINA neutronic module of the LEGO code for steady-state and transient PWR plant simulations

International Nuclear Information System (INIS)

Nicolopoulos, D.; Pollacchini, L.; Vimercati, G.; Spelta, S.

1989-01-01

The Automation Research Center (CRA) of ENEl has implemented some models for analyzing both incidental and operational transients in PWR power plants. For such models an axial neutron kinetics module characterized by high computational efficency with adequate results accuracy was called for. CISE has been entrusted with the task of implementing such a module named KINA and based on IQS (Improved Quasi Static) method, to be included in the library of LEGO modular code used by CRA to set up PWR power models. Moreover, The KINA module has been adapted to the neutron constants computing model developed by the EdF-SEPTEN, which has been using and improving the LEGO code for a long time in cooperation with ENEL-CRA. In this paper, after some remarks on the LEGO code, a general description of KINA neutronic module is given. The resylts of a preliminary validation activity of KINA for an EdF 1300 MWe PWR plant are also presented

EDF decommissioning and dismantling policy a global commitment to safety, environment and cost efficiency of nuclear energy

International Nuclear Information System (INIS)

Rondeau, J.

2001-01-01

Until recently, EDF's policy regarding the dismantling of its decommissioned nuclear power plants was to reach 'level 2' (release of non-nuclear facilities) and to postpone final dismantling for another 30-40 years. Today, some studies suggest that a full deconstruction program of the first generation NPPs (9 units) could be optimized over the period 2000 - 2025. EDF has acquired during the last ten years an unique experience, both as an operator and as an engineering company, in the frame of the decommissioning programme of its own NPPs. Many types of reactors, including graphite moderated one, PWR, are at varying stages of the dismantling process.Plant operation quality is at the core of a satisfactory control of releases. Over the last decade, as a result of the efforts of all operating sites associated with good in-house operating practice feedback, the overall release volume has been divided by two, and the release activity by one hundred. Another issue given increased attention is radiological cleanliness. EDF-DPN launched a 'radiological cleanliness' action plan revolving around two main themes: increased monitoring of nuclear-related transportations, site entrance and access to controlled areas, along with on-site radiological cleanliness, particularly during maintenance work tasks. Progress is already apparent in several points at issue and the overall objective of the action plan should be attained. (author)

Temperature escalation in PWR fuel rod simulators due to the zircaloy/steam reaction: Tests ESSI-1,2,3

International Nuclear Information System (INIS)

Hagen, S.; Malauschek, H.; Wallenfels, K.P.; Peck, S.O.

1983-08-01

This report discusses the test conduct, results, and posttest appearance of three scoping tests (ESSI-1,2,3) investigating temperature escalation in zircaloy clad fuel rods. The experiments are part of an out-of-pile program using electrically heated fuel rod simulators to investigate PWR fuel element behavior up to temperatures of 2000 0 C. These experiments are part of the PNS Severe Fuel Damage Program. The temperature escalation is caused by the exothermal zircaloy/steam reaction, whose reaction rate increases exponentially with the temperature. The tests were performed using different initial oxide layers as a major parameter, obtained by varying the heatup rates and steam exposure times. (orig./RW) [de

Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

Energy Technology Data Exchange (ETDEWEB)

Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

2016-08-01

To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

State of the art report of exponential experiments with PWR spent nuclear fuel

International Nuclear Information System (INIS)

Ro, Seung Gy; Park, Sung Won; Park, Kwang Joon; Kim, Jong Hoon; Hong, Kwon Pyo; Shin, Hee Sung

2000-09-01

Exponential experiment method is discussed for verifying the computer code system of the nuclear criticality analysis which makes it possible to apply for the burnup credit in storage, transportation, and handling of spent nuclear fuel. In this report, it is described that the neutron flux density distribution in the exponential experiment system which consists of a PWR spent fuel in a water pool is measured by using 252 Cf neutron source and a mini-fission chamber, and therefrom the exponential decay coefficient is determined. Besides, described is a method for determining the absolute thermal neutron flux density by means of the Cd cut-off technique in association with a gold foil. Also a method is described for analyzing the energy distribution of γ-ray from the gold foil activation detector in detail

Cost comparisons of wet and dry interim storage facilities for PWR spent nuclear fuel in Korea

International Nuclear Information System (INIS)

Cho, Chun-Hyung; Kim, Tae-Man; Seong, Ki-Yeoul; Kim, Hyung-Jin; Yoon, Jeong-Hyoun

2011-01-01

Research highlights: → We compare the costs of wet and dry interim storage facilities for PWR spent fuel. → We use the parametric method and quotations to deduce unknown cost items. → Net present values and levelized unit prices are calculated for cost comparisons. → A system price is the most decisive factor in cost comparisons. - Abstract: As a part of an effort to determine the ideal storage solution for pressurized water reactor (PWR) spent nuclear fuel, a cost assessment was performed to better quantify the competitiveness of several storage types. Several storage solutions were chosen for comparison, including three dry storage concepts and a wet storage concept. The net present value (NPV) and the levelized unit cost (LUC) of each solution were calculated, taking into consideration established scenarios and facility size. Wet storage was calculated to be the most expensive solution for a 1700 MTU facility, and metal cask storage marked the highest cost for a 5000 MTU facility. Sensitivity analyses on discount rate, metal cask price, operation and maintenance cost, and facility size revealed that the system price is the most decisive factor affecting competitiveness among the storage types.

Cost comparisons of wet and dry interim storage facilities for PWR spent nuclear fuel in Korea

Energy Technology Data Exchange (ETDEWEB)

Cho, Chun-Hyung, E-mail: skycho@krmc.or.kr [Korea Radioactive Waste Management Corporation, 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Kim, Tae-Man; Seong, Ki-Yeoul; Kim, Hyung-Jin; Yoon, Jeong-Hyoun [Korea Radioactive Waste Management Corporation, 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of)

2011-05-15

Research highlights: > We compare the costs of wet and dry interim storage facilities for PWR spent fuel. > We use the parametric method and quotations to deduce unknown cost items. > Net present values and levelized unit prices are calculated for cost comparisons. > A system price is the most decisive factor in cost comparisons. - Abstract: As a part of an effort to determine the ideal storage solution for pressurized water reactor (PWR) spent nuclear fuel, a cost assessment was performed to better quantify the competitiveness of several storage types. Several storage solutions were chosen for comparison, including three dry storage concepts and a wet storage concept. The net present value (NPV) and the levelized unit cost (LUC) of each solution were calculated, taking into consideration established scenarios and facility size. Wet storage was calculated to be the most expensive solution for a 1700 MTU facility, and metal cask storage marked the highest cost for a 5000 MTU facility. Sensitivity analyses on discount rate, metal cask price, operation and maintenance cost, and facility size revealed that the system price is the most decisive factor affecting competitiveness among the storage types.

VAMCIS, a new measuring channel for continuous monitoring of leak rates inside PWR steam generators

International Nuclear Information System (INIS)

Champion, G.; Dubail, A.; Lefevre, F.

1988-01-01

In order to assess the primary to secondary leakage, radioactive isotopes, formed in the primary coolant as a result of fission or neutron capture, are usually monitored in the pressurized water reactor (PWR) secondary coolant. Conventional methods mainly based on the detection of 133 Xe, tritium, and 41 Ar are widely used on French Electricite de France (EdF) PWRs. Some years ago, it appeared necessary to improve both leak rate assessments and steam generator tube rupture (SGTR) detection. A volumetric activity measuring channel inside steam (VAMCIS) has been developed for this purpose. The SGTR that occurred at the North Anna PWR has focused the attention of safety authorities on this new measuring channel. It is planned to implement VAMCIS at North Anna in order to check the leak rate variations more accurately

Fuel performance under normal PWR conditions: A review of relevant experimental results and models

Science.gov (United States)

Charles, M.; Lemaignan, C.

1992-06-01

Experiments conducted at Grenoble (CEA/DRN) over the past 20 years in the field of nuclear fuel behaviour are reviewed. Of particular concern is the need to achieve a comprehensive understanding of and subsequently overcome the limitations associated with high burnup and load-following conditions (pellet-cladding interaction (PCI), fission gas release (FGR), water-side corrosion). A general view is given of the organization of research work as well as some experimental details (irradiation, postirradiation examination — PIE). Based on various experimental programmes (Cyrano, Medicis, Anemone, Furet, Tango, Contact, Cansar, Hatac, Flog, Decor), the main contributions of the thermomechanical behaviour of a PWR fuel rod are described: thermal conductivity, in-pile densification, swelling, fission gas release in steady state and moderate transient conditions, gap thermal conductance, formation of primary and secondary ridges under PCI conditions. Specific programmes (Gdgrif, Thermox, Grimox) are devoted to the behaviour of particular fuels (gadolinia-bearing fuel, MOX fuel). Moreover, microstructure-based studies have been undertaken on fission gas release (fine analysis of the bubble population inside irradiated fuel samples), and on cladding behaviour (PCI related studies on stress-corrosion cracking (SCO, irradiation effects on zircaloy microstructure).

A review on the heterogeneous thorium fuel concept for PWR applications

International Nuclear Information System (INIS)

Joo, H. K.; Noh, J. M.; Yoo, J. W.; Kim, K. H.

2001-08-01

Seed-blanket unit (SBU) and whole assembly seed and blanket (WASB) are being investigated for the PWR application as well as homogeneous thorium fuel under the US NERI program. For the verification of HELIOS capability for thorium analysis, the characteristics of heterogeneous thorium fuels was evaluated by HELIOS color-set calculation and compared with the calculation results of the US NERI. The infinite multiplication factors from HELIOS calculation are in good agreement with CASMO-4 except for SBU which uses metallic fuel for seed material. The maximum relative difference in power distribution is occurred in WASB case, and is about 5% compared to MCNP. The isotopic concentrations for Am-241, Am-243, and Cm-244 of HELIOS agree well with CASMO-4's, but show a significant discrepancy from MOCUP mainly caused by the old data of cross section and decay constants in ORIGEN. The nonproliferation characteristic of thorium-based fuel such as critical mass, spontaneous fission rate, decay heat generation rate are superior to the conventional uranium fuel. Even though the diversion of U-233 produced in blanket is a technically difficult, the enrichment of uranium isotopes including U-233 is slightly over the limit for safeguard aspects. The urnaium contents in thorium fuel is need to be adjusted in order to meet the safeguard limit. A preliminary assessment of fuel economics was performed based on the uranium utilization and SWU utilization. The natural uranium utilization factors of heterogeneous thorium-based fuel increased by 10δ18%, but the SWU utilization factor decreased by 6-δ11% compared to uranium fuel. The cost of uranium purchase of 50USI/KgU and SWU cost of 110USI/SWU-Kg, recommended by OECD/NEA, gives a comparable economics of thorium-based fuel to uraium fuel. The detailed fuel cycle analysis will take account of the other factors like the variation of uranium purchase cost and SWU cost, fabrication cost of thorium fuel, thorium purchase cost, the capcity

A review on the heterogeneous thorium fuel concept for PWR applications

Energy Technology Data Exchange (ETDEWEB)

Joo, H. K.; Noh, J. M.; Yoo, J. W.; Kim, K. H

2001-08-01

Seed-blanket unit (SBU) and whole assembly seed and blanket (WASB) are being investigated for the PWR application as well as homogeneous thorium fuel under the US NERI program. For the verification of HELIOS capability for thorium analysis, the characteristics of heterogeneous thorium fuels was evaluated by HELIOS color-set calculation and compared with the calculation results of the US NERI. The infinite multiplication factors from HELIOS calculation are in good agreement with CASMO-4 except for SBU which uses metallic fuel for seed material. The maximum relative difference in power distribution is occurred in WASB case, and is about 5% compared to MCNP. The isotopic concentrations for Am-241, Am-243, and Cm-244 of HELIOS agree well with CASMO-4's, but show a significant discrepancy from MOCUP mainly caused by the old data of cross section and decay constants in ORIGEN. The nonproliferation characteristic of thorium-based fuel such as critical mass, spontaneous fission rate, decay heat generation rate are superior to the conventional uranium fuel. Even though the diversion of U-233 produced in blanket is a technically difficult, the enrichment of uranium isotopes including U-233 is slightly over the limit for safeguard aspects. The urnaium contents in thorium fuel is need to be adjusted in order to meet the safeguard limit. A preliminary assessment of fuel economics was performed based on the uranium utilization and SWU utilization. The natural uranium utilization factors of heterogeneous thorium-based fuel increased by 10{delta}18%, but the SWU utilization factor decreased by 6-{delta}11% compared to uranium fuel. The cost of uranium purchase of 50USI/KgU and SWU cost of 110USI/SWU-Kg, recommended by OECD/NEA, gives a comparable economics of thorium-based fuel to uraium fuel. The detailed fuel cycle analysis will take account of the other factors like the variation of uranium purchase cost and SWU cost, fabrication cost of thorium fuel, thorium purchase cost

Forced transitions for EDF: the EDF ship facing head-winds; Dalkia, the renewable trump card to be valorized

International Nuclear Information System (INIS)

Barbaux, Aurelie

2018-01-01

Electricite de France (EDF), the state-owned and main power producer in France is, as the Titanic liner, lacking agility to navigate into the wild seas of the French energy transition policy. Besides, the era of engineers that construct energy cathedrals and of nuclear full supremacy is now well over, giving the way to innovative management policies with workers participation and digitalisation. EDF is preparing a new strategy addressing market opening and deregulation, renewable energy sources, and more freedom for EDF's numerous subsidiary companies. A second paper precisely concerns Dalkia, one of EDF's subsidiaries which is specialized in energy efficiency, heat and cold generation and networks that are more and more supplied by renewable energies. A third paper addresses EDF-SEI, whose specialty is energy autonomy in isolated areas (islands, etc.)

A study on the thermal expansion characteristics of simulated spent fuel and simulated DUPIC fuel

International Nuclear Information System (INIS)

Kang, Kweon Ho; Ryu, H. J.; Kim, H. S.; Song, K. C.; Yang, M. S.

2001-10-01

Thermal expansions of simulated spent PWR fuel and simulated DUPIC fuel were studied using a dilatometer in the temperature range from 298 to 1900 K. The densities of simulated spent PWR fuel and simulated DUPIC fuel used in the measurement were 10.28 g/cm3 (95.35 % of TD) and 10.26 g/cm3 (95.14 % of TD), respectively. Their linear thermal expansions of simulated fuels are higher than that of UO2, and the difference between these fuels and UO2 increases progressively as temperature increases. However, the difference between simulated spent PWR fuel and simulated DUPIC fuel can hardly be observed. For the temperature range from 298 to 1900 K, the values of the average linear thermal expansion coefficients for simulated spent PWR fuel and simulated DUPIC fuel are 1.391 10-5 and 1.393 10-5 K-1, respectively. As temperature increases to 1900 K, the relative densities of simulated spent PWR fuel and simulated DUPIC fuel decrease to 93.81 and 93.76 % of initial densities at 298 K, respectively

A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

International Nuclear Information System (INIS)

Hiroyuki Yoshida; Akira Ohnuki; Hajime Akimoto

2002-01-01

A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation

EDF's fiftieth anniversary

International Nuclear Information System (INIS)

Spann, L.

1996-01-01

During its half-century of existence, EdF (Electricite de France) has carried out an impressive productivity effort. The company, the manpower of which has remained practically the same from the beginning, produces today 22 times more kW per hour than in 1946. In 1995, the French production of electric power reached 442.2 TWh, which corresponds to a 3.9 % increase with respect to the previous year. This production corresponds to a 99000 MW installed capacity of nuclear (59 %), thermal (17 %) and hydraulic (24 %) origin, respectively. This example illustrates the success of EdF, which is considered as ''the result of a huge technical and industrial adventure'' said Franck Borotra, French Ministry of Industry, Post and Telecommunications. This paper comprises three parts. The first part describes some historical and political aspects of the creation and development of the company with the building of several hydro-electric dams up to 1957, the starting of the nuclear program in 1973-74 during the petroleum crisis, the new goals and the evolution of the company inside the European Community. The second part describes the political, ecological and financial controversy about the Saone-Rhine rivers connection project, in which EdF is involved, and which would allow a 15 times increase of the rivers transportation capacity. The last part is a comparative summary of EdF's 1994 and 1995 balance sheet of production, financing, expenditures, trade, contracts etc.. (J.S.). 2 photos

Pressure loss tests for DR-BEP of fullsize 17 x 17 PWR fuel assembly

International Nuclear Information System (INIS)

Chung, Moon Ki; Chun, Se Young; Chang, Seok Kyu; Won, Soon Youn; Cho, Young Rho; Kim, Bok Deuk; Min, Kyoung Ho

1993-01-01

This report describes the conditions, procedure and results in the pressure loss tests carried out for a double grid type debris resistance bottom end piece (DR-BEP) designed by KAERI. In this test, the pressure loss coefficients of the full size 17 x 17 PWR simulated fuel assembly with DR-BET and with standard-BEP were measured respectively, and the pressure loss coefficients of DR-BEP were compared with the coefficients of STD-BET. The test conditions fall within the ranges of loop pressure from 5.2 to 45 bar, loop temperature from 27 to 221 deg C and Reynolds number in fuel bundle from 2.17 x 10 4 to 3.85 x 10 5 . (Author) 5 refs., 18 figs., 5 tabs

A study on the direct use of spent PWR fuel in CANDU reactors -Fuel management and safety analysis-

Energy Technology Data Exchange (ETDEWEB)

Park, Hyun Soo; Lee, Boh Wook; Choi, Hang Bok; Lee, Yung Wook; Cho, Jae Sun; Huh, Chang Wook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

The reference DUPIC fuel composition was determined based on the reactor safety, thermal-hydraulics, economics, and refabrication aspects. The center pin of the reference DUPIC fuel bundle is poisoned with natural dysprosium. The worst LOCA analysis has shown that the transient power and heat deposition of the reference DUPIC core are the same as those of natural uranium CANDU core. The intra-code comparison has shown that the accuracy of DUPIC physics code system is comparable to the current CANDU core design code system. The sensitivity studies were performed for the refuelling schemes of DUPIC core and the 2-bundle shift refuelling scheme was selected as the standard refuelling scheme of the DUPIC core. The application of 4-bundle shift refuelling scheme will be studied in parallel as the auto-refuelling method is improved and the reference core parameters of the heterogeneous DUPIC core are defined. The heterogeneity effect was analyzed in a preliminary fashion using 33 fuel types and the random loading strategy. The refuelling simulation has shown that the DUPIC core satisfies the current CANDU 6 operating limits of channel and bundle power regardless of the fuel composition heterogeneity. The 33 fuel types used in the heterogeneity analysis was determined based on the initial enrichment and discharge burnup of the PWR fuel. 90 figs, 62 tabs, 63 refs. (Author).

2002 annual report EDF group

International Nuclear Information System (INIS)

2002-01-01

This document is the 2002 annual report of Electricite de France (EdF) group, the French electric utility. Content: Introductory section (EDF at a glance, Chairman's message, 2002 Highlights); Corporate governance and Group strategy (Corporate governance, sustainable growth strategy, EDF branches); Financial performance (Reaching critical mass, Margins holding up well, Balance sheet); Human resources (Launching Group-wide synergies, Optimising human resources); Customers (Major customers, SMEs and professional customers, Local authorities, Residential customers, Ensuring quality access to electricity); Generation (A balanced energy mix, Nuclear generation, Fossil-fuelled generation, Renewable energies); Corporate social responsibility (Global and local partnerships, Promoting community development)

EDF Group - Annual Report 2010

International Nuclear Information System (INIS)

2011-01-01

The EDF Group is one of the world's leading energy companies, active in all areas from generation to trading and network management. It has a sound business model, evenly balanced between regulated and deregulated activities. With its first-rate human resources, R and D capability, expertise in engineering and operating generation plants and networks, as well as its energy eco-efficiency offers, the Group delivers competitive solutions that help ensure sustainable economic development and climate protection. The EDF Group is the leader in the French and UK electricity markets and has solid positions in Italy and numerous other European countries, as well as industrial operations in Asia and the United States. Everywhere it operates, the Group is a model of quality public service for the energy sector. This document is EDF Group's annual report for the year 2010. It contains information about Group profile, governance, business, development strategy, sales and marketing, positions in Europe and international activities. The document is made of several reports: the Activity and Sustainable Development Report, the Financial Report, the Management Report, the Report by the Chairman of EDF Board of Directors on corporate governance and internal control procedures, the Milestones report, the 'EDF at a glance' report, and the Sustainable Development Indicators

Transition from uranium to denatured uranium/thorium fuel in an existing PWR

International Nuclear Information System (INIS)

Walters, M.A.

1982-01-01

The purpose of this research was to determine whether it is possible to make a gradual transition from uranium to denatured uranium/thorium (DUTH) fuel in an existing PWR by adding DUTH assemblies during each scheduled refueling and, if the transition is possible, to develop a general procedure for making it. The feasibility of the transition was established by identifying acceptable refueling schemes for a series of transition cores, and in the process, a method for identifying acceptable schemes evolved. The utility of the method was then demonstrated by applying it to a standard reactor operating under normal conditions. The vehicle used to examine proposed fuel mixtures and to select acceptable ones was a set of one-dimensional computer codes. The core was modeled as a set of five concentric fuel zones with a reflector. Fuel mixtures were proposed and the computer codes were used to determine whether a mixture was acceptable, i.e., whether it had the desired k-effective and flux and power distributions. The parameters allowed to vary in selection of proposed fuel mixtures were enrichment of fresh fuel assemblies, number of uranium and DUTH assemblies added during each refueling, and distribution of fuel in the core. Results of the research showed that a gradual transition is possible. Furthermore, there is a method that allows the identification of fuel mixtures that are likely to be acceptable. It requires the calculation of K-infinity for the entire proposed core and for some of its regions. These values of K-infinity and relationships developed in this research can be used to predict the flux distribution and the final k-effective for the proposed fuel mixture

Chemical cleaning of PWR steam generators: application at Nogent 1

International Nuclear Information System (INIS)

Fiquet, J.M.; Veysset, J.P.; Esteban, L.; Saurin, P.

1990-01-01

EDF has developed and patented a chemical cleaning process for PWR steam generators, based on the use of a mixture of organic acids in order to: - dissolve iron oxides and copper with a single solution; - clean dented crevices. Qualification tests have permitted to demonstrate effectiveness of the solution and its inocuousness related to steam generator materials. The process, the license of which belongs to SOMAFER R.A. and FRAMATOME, has been implemented in France at Nogent. The goal was to dissolve iron oxides allowing metallic particles, aggregated on the tubesheet, to be released and mechanically removed. The effectiveness was satisfactory and this treatment is to be extended to other units [fr

Optimization of the distribution of bars with gadolinium oxide in reactor fuel elements PWR; Optimizacion de la distribucion de barras con oxido de gadolinio en elementos combustibles para reactores PWR

Energy Technology Data Exchange (ETDEWEB)

Melgar Santa Cecilia, P. A.; Velazquez, J.; Ahnert Iglesias, C.

2014-07-01

In the schemes of low leakage, currently used in the majority of PWR reactors, it makes use of absorbent consumables for the effective control of the factors of peak, the critical concentration of initial boron and the moderator temperature coefficient. One of the most used absorbing is the oxide of gadolinium, which is integrated within the fuel pickup. Occurs a process of optimization of fuel elements with oxide of gadolinium, which allows for a smaller number of configurations with a low peak factor for bar. (Author)

The deformation of PWR fuel in a LOCA

International Nuclear Information System (INIS)

Mann, C.A.; Hindle, E.D.; Parsons, P.D.

1982-04-01

Available world-wide published data on the deformation of PWR fuel in a loss-of-coolant accident are reviewed. Adequate data exist for the oxidation of Zircaloy up to about 1500 0 C; data are increasingly sparse above this temperature and lacking above the melting point. The US NRC criteria for embrittlement are discussed and considered adequate for undeformed cladding, though they may be less so for deformed thinned material. Cladding deformation and the factors controlling it are considered in the light of data from the US, Germany, Japan and the UK. It is concluded that strains in the range 30% - 70% can be produced in experiments simulating LOCA conditions. The behaviour of cladding is strongly influenced by the spatial distribution of temperature, which is in turn dependent on heat transfer mechanisms at the surfaces of the cladding. No realistic experiment, i.e. one with a multirod array and simulated cooling, has produced deformations which would inhibit quenching. Such experiments have not, however, as yet covered the entire range of conditions which might obtain following a LOCA. (author)

EDF - The Inspector General's report on Nuclear Safety and Radiation Protection 2009

International Nuclear Information System (INIS)

2010-01-01

After a first chapter in which the Inspector General states his own vision of facts and results regarding nuclear safety and radiation protection for 2009, the relationship with the ASN (Nuclear safety authority), and also gives his own impression on the whole EDF nuclear system, the next chapters more particularly address the following topics: safety management, radiation protection of interveners, abilities and education for new challenges, the exploitation quality, the implementation of new regulatory arrangements, the extension of exploitation duration, the EPR project, the doctors as actors of safety management, service providing or subcontracting companies and fleet performance, prevention and struggle against fire, the activity of British Energy which is part of the EDF group, events which occurred within the EDF group (exposure of an intervener, incidents involving fuel components, biases in work monitoring, a fire on a turbo generator, an incident in Dungeness), lessons drawn from other industrial events (in a Russian hydroelectric station and in a refinery in Texas), and visits in China and in Japan

Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

Energy Technology Data Exchange (ETDEWEB)

LaFleur, Adrienne M., E-mail: alafleur@lanl.gov; Menlove, Howard O., E-mail: hmenlove@lanl.gov

2015-05-01

Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. - Highlights: • Experimental measurements of PWR fresh and spent FAs were performed with SINRD. • Good agreement of MCNPX and measured results confirmed accuracy of SINRD model. • For fresh fuel, SINRD and PNMC ratios were not sensitive to water gaps of ≤5-mm. • Practical use of SINRD would be in Fork detector to reduce systematic uncertainties.

Nuclear regulatory guides for LWR (PWR) fuel in Japan and some related safety research

International Nuclear Information System (INIS)

Ichikawa, M.

1994-01-01

The general aspects of licensing procedure for NPPs in Japan and regulatory guides are described. The expert committee reports closely related to PWR fuel are reviewed. Some major results of reactor safety research experiments at NSPR (Nuclear Safety Research Reactor of JAERI) used for establishment of related guide, are discussed. It is pointed out that the reactor safety research in Japan supports the regularity activities by establishing and revising guides and preparing the necessary regulatory data as well as improving nuclear safety. 10 figs., 4 refs

Nuclear regulatory guides for LWR (PWR) fuel in Japan and some related safety research

Energy Technology Data Exchange (ETDEWEB)

Ichikawa, M [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

1994-12-31

The general aspects of licensing procedure for NPPs in Japan and regulatory guides are described. The expert committee reports closely related to PWR fuel are reviewed. Some major results of reactor safety research experiments at NSPR (Nuclear Safety Research Reactor of JAERI) used for establishment of related guide, are discussed. It is pointed out that the reactor safety research in Japan supports the regularity activities by establishing and revising guides and preparing the necessary regulatory data as well as improving nuclear safety. 10 figs., 4 refs.

Main prospects of EDF's nuclear program

International Nuclear Information System (INIS)

Pierre Bacher, M.; Jean Pierrard, M.

1994-01-01

Today, EDF is at a half way point in its third major standardized series, the N4 1400 MW series. The main objective agreed upon for this N4 series, was to improve the insertion of man in the control loop. After the TMI accident in 1979, selective improvements had already been introduced in the 900 series ; the 1300 series, which was the underway, was also the object of more significant improvements (for instance a digital control system) ; but it still seemed desirable that all the lessons to be learned from the accident be reflected in the third series : the fully computerized instrumentation and control system developed for the N4 series ?comparable to the one developed for Airbus ?is today the first of its kind in the world. The demand increase for electric energy dramatically stopped in 1993. However, the prospect for the coming years is more optimistic. Electricity consumption is expected to increase slowly by an average 2% per year. In this context, EDF is preparing its energy program for the nest century and takes into account that hydraulic generation capacity will remain stable and that the French thermal units fueled with coal are rather old and will have to be replaced.

EDF decommissioning programme a global commitment to safety, environment and cost efficiency of nuclear energy

International Nuclear Information System (INIS)

Grenouillet, J.-J.

2002-01-01

EDF has 9 NPPs permanently shutdown and under decommissioning. EDF considers that if the nuclear option is to remain open, it is necessary to deal with increasing public concerns for environmental and waste management issues. Therefore EDF has decided to achieve total dismantling of all shutdown reactor in the next 25 years. The Decommissioning Program has been developed including 2 stages of activities. The first stage consists of: 1) Final dismantling of Brennilis in 2015; 2) A dismantling demonstration of a PWR reactor building (Chooz A) before starting replacing the population of PWRs currently in operation; 3) Final dismantling of reactor containment of a GCR (Bugey 1) as a first of its kind. The second stage includes: 1)Dismantling of following 5 GCR (Saint Laurent A1 and A2, Chinon A1, A2 and A3); 2) Final dismantling of Chooz A and Bugey 1 in 2025. The successful implementation relies on the simplification of the regulatory process; availability of treatment, conditioning and disposal facilities and effective nuclear industry. The main issue is availability of time and waste solutions such as opening of a Very Low Waste disposal in 2003 (130 000 tons); opening of a new disposal for graphite and radiferous wastes (17 000 tons) in 2010 and opening in 2007-2008 of a centralized interim storage (BANEDA) facility for long-lived Medium Level Wastes (500 tons including filters, control rods etc)Three investigations are to be carried out for high level radioactive waste before 2006

Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design

International Nuclear Information System (INIS)

Fernandez, J. L.; Lopez, J.

1986-01-01

Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs

VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

Directory of Open Access Journals (Sweden)

Muhammad Darwis Isnaini

2016-03-01

Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

Spent fuel and fuel pool component integrity. Annual report, FY 1979

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

1980-05-01

International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-μm) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report

Advantages for EDF of using and updating PSAs for the probabilistic analysis of accident scenarios in nuclear power plants

International Nuclear Information System (INIS)

Feuillade, Gilles

2000-01-01

This paper shows the advantages for EDF of using and updating PSA models of PWR units for the probabilistic assessment of accident scenarios. These advantages may be classified in various categories: The construction of PSA models makes it possible to aggregate knowledge in a variety of fields: thermohydraulics, the behavior of equipment and systems, organization, plant operation, etc. The updating of PSA models makes it possible to monitor the state of progress of PWR unit safety levels and also allows a variety of applications to be used throughout the service life of the unit. The results obtained are directly applicable to the units as the 'reference PSA models' developed by EDF conform to the units in service. The use of PSA for the examination of incident or accident scenarios makes it possible to verify the adequacy of the resources both in terms of 'systems' and 'plant operation'. These two notions are to be taken in the broadest sense, as they cover the aspects of system design, reliability and availability of equipment, organization of plant operation, comprehensiveness of operating procedures, human redundancy, etc. The use of PSA models through the different applications (analyses of predominant sequences, analyses of main equipment failures, analyses main operator actions, analyses according to the power units, etc) is becoming an indispensable supplement to conventional deterministic analyses of the envisaged accident scenarios. Within the scope of accident prevention, it constitutes a tool to assist the decision-maker, especially when evaluating the pertinence of the General Operating Rules (operating procedures, operating technical specifications, periodic testing, etc). (S.Y.)

Analysis of transient heat conduction in a PWR fuel rod by an improved lumped parameter approach

Energy Technology Data Exchange (ETDEWEB)

Dourado, Eneida Regina G. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Cotta, Renato M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Mecanica; Jian, Su, E-mail: eneidadourado@gmail.com, E-mail: sujian@nuclear.ufrj.br, E-mail: cotta@mecanica.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2017-07-01

This paper aims to analyze transient heat conduction in a nuclear fuel rod by an improved lumped parameter approach. One-dimensional transient heat conduction is considered, with the circumferential symmetry assumed and the axial conduction neglected. The thermal conductivity and specific heat in the fuel pellet are considered temperature dependent, while the thermophysical properties of the cladding are considered constant. Hermite approximation for integration is used to obtain the average temperature and heat flux in the radial direction. Significant improvement over the classical lumped parameter formulation has been achieved. The proposed model can be also used in dynamic analysis of PWR and nuclear power plant simulators. (author)

Analysis of transient heat conduction in a PWR fuel rod by an improved lumped parameter approach

International Nuclear Information System (INIS)

Dourado, Eneida Regina G.; Cotta, Renato M.; Jian, Su

2017-01-01

This paper aims to analyze transient heat conduction in a nuclear fuel rod by an improved lumped parameter approach. One-dimensional transient heat conduction is considered, with the circumferential symmetry assumed and the axial conduction neglected. The thermal conductivity and specific heat in the fuel pellet are considered temperature dependent, while the thermophysical properties of the cladding are considered constant. Hermite approximation for integration is used to obtain the average temperature and heat flux in the radial direction. Significant improvement over the classical lumped parameter formulation has been achieved. The proposed model can be also used in dynamic analysis of PWR and nuclear power plant simulators. (author)

A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

Energy Technology Data Exchange (ETDEWEB)

Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

EDF Group: activity and sustainable development 2010

International Nuclear Information System (INIS)

2010-01-01

After a map illustrating the worldwide dimension of EDF, and an interview of the EDF's chairman, this report presents and comments some sustainable development indicators for EDF activity in 2010. Then, after a presentation of the world energy context, it presents and comments EDF's main activities: energy production and engineering, network management, trading. It presents and comments the operational performance of the company in different countries (France, United Kingdom, Italy, and other countries) and evokes additional activities. The next chapter deals with human resources and with innovation and R and D activities. The last chapter presents governance structures and financial results

EDF Energies Nouvelles. Financial report at June 30, 2011

International Nuclear Information System (INIS)

2011-01-01

EDF Energies Nouvelles is a world leader in renewable energy electricity. The company develops, builds and operates clean energy power plants both for its own account and for third parties. Historically, EDF Energies Nouvelles primarily developed its business in two geographical areas, Europe and North America (U.S., Canada and Mexico). EDF Energies Nouvelles is a subsidiary of EDF, helping the Group to achieve its renewable energy goals. The EDF Group generates low-carbon electricity around the world and actively participates in the energy transition. EDF Energies Nouvelles prioritizes development of wind and photovoltaic solar capacity. As an integrated operator with global reach, EDF Energies Nouvelles covers the entire renewable energy chain, from development to operation and maintenance, and manages all project phases in-house. This document is EDF Energies Nouvelles's half-year financial report for 2011. It contains a half-year activity report, the consolidated financial statements at June 30, 2011 and the report drafted by the Statutory Auditors

Chemistry evaluation in French EDF Nuclear Power Plants

International Nuclear Information System (INIS)

Jacquier, Hervé

2014-01-01

The Nuclear Production Division of EDF is comprised of 19 power stations (58 PWR reactors) and 2 national engineering organisations. Nuclear Inspection (IN) is an internal assessment unit of the EDF Nuclear Production Directorate. At the request of the Directorate, it carries out periodic evaluations of all the units of the division. The evaluation of the nuclear sites (EGE: Overall Excellence Assessment) is carried out every 4 years, an intermediate evaluation is also carried out between each EGE. These evaluations are independent of the WANO and IAEA evaluations. Exchanges are carried out between Nuclear Inspection and the other international operators (for example, USA (INPO), England, China...) to share site evaluation methods. These evaluations are carried out by a team of 30 inspectors, reinforced during each evaluation by 10 peers who come from the various French nuclear sites. Nuclear Inspection produces a performance standards document for each FUNCTIONAL AREA, which is based on the requirements of the company. On the whole, 13 areas are evaluated during each inspection, in particular: Management, Operations, Maintenance, Engineering and Chemistry. The area of reactor plant chemistry has been evaluated since 2009. The Chemistry performance standards document is written from the EDF internal requirements and international references. During site evaluations, all the performance standards are assessed for compliance. The Chemistry performance standards document is comprised of 3 topics: Management of plant chemistry, The respect of the chemical and radiochemical specifications, The condition of the laboratories and the sampling lines, measuring equipment, and chemical products. The evaluations carried out make it possible to define strengths and weaknesses which the sites must address. After each evaluation, the assessment is presented to the site management and to the director of EDF Nuclear Production. For 4 years these evaluations have allowed progress to

Dynamic structural analysis for assemblies of fuel elements in the core of a PWR

International Nuclear Information System (INIS)

Silva Macedo, L.V. da.

1991-01-01

It is presented a procedure for the dynamic structural analysis of a PWR core. Impacts between fuel assemblies may occur because of the existence of gaps between them. Thus, the problem is non-linear and an spectral analysis is avoided. It is necessary a time-history response analysis. The Modal Superposition Method with the Duhamel integral was used in order to solve the problem. It is presented an algorithm of solution and also results obtained with the STYCA computer program, developed in the basis of what was proposed here. (author)

Reactivity Impact of Difference of Nuclear Data Library for PWR Fuel Assembly Calculation by Using AEGIS Code

International Nuclear Information System (INIS)

Ohoka, Yasunori; Tatsumi, Masahiro; Sugimura, Naoki; Tabuchi, Masato

2011-01-01

In 2010, the latest version of the Japanese Evaluated Nuclear Data Library (JENDL-4.0) has been released by JAEA. JENDL-4.0 is major update from JENDL- 3.3, and confirmed to give good accuracy by integral test for fission reactor systems such as fast neutron system and thermal neutron system. In this study, we evaluated the reactivity impact due to difference between ENDF/B-VII.0 and JENDL-4.0 for PWR fuel assembly burnup calculation using AEGIS code which has been developed by Nuclear Engineering, Ltd. in cooperation with Nuclear Fuel Industries, Ltd. and Nagoya University

Sizewell: proposed site for Britain's first PWR power station

International Nuclear Information System (INIS)

1980-10-01

The pamphlet covers the following points, very briefly: nuclear power - a success story; the Government's nuclear programme; why Sizewell; the PWR (with diagram); the PWR at Sizewell (with aerial view) (location; size; cooling water; road access; fuel transport; construction; employment; environment; screening; the next steps (licensing procedures, etc.); safety; further information). (U.K.)

The French approach through the SMR consortium (CEA, EDF, AREVA, DCNS)

International Nuclear Information System (INIS)

Chenais, J.; Diet, A.; Grondin, Y.; Perrier, S.

2014-01-01

The SMR (Small Modular Reactor) is a new range of nuclear reactors characterized by a low output power (less than 300 MWe) and a standardized modular factory-based construction. Moderate capital costs make this type of reactor very attractive for medium-size power utilities or small countries. A consortium made around the CEA and its industrial partners (EDF, AREVA and DCNS) has performed economic and technical feasibility studies. It appears that the competitiveness of SMR require power production costs around 100 euros/MWh and that SMR will be based on PWR technology with a power output of 150 MWe per unit. 2 types of SMR are proposed: a terrestrial unit and an immersed unit. The concept of the immersed SMR, called Flexblue, is based on an unlimited access to a cold sink (the ocean) and a complete modularity. In both types the standard of safety appears to be high. (A.C.)

EDF Group - Annual Report 2016

International Nuclear Information System (INIS)

2017-01-01

EDF group is the world's leading electricity company and global leader for low-carbon energy production. Particularly well established in Europe, especially France, the United-Kingdom, Italy and Belgium, as well as North and South America, the Group covers all businesses spanning the electricity value chain - from generation to distribution and including energy transmission and trading activities - to continuously balance supply. A marked increase in the use of renewables is bringing change to its electricity generation operations, which are underpinned by a diversified and complementary energy mix founded on nuclear power capacity. EDF offers products and advice to help residential customers manage their electricity consumption, to support the energy and financial performance of its business customers, and to help local authorities find sustainable solutions. This document is EDF Group's annual report for the year 2016. It contains information about Group profile, governance, business, development strategy, sales and marketing, positions in Europe and international activities. The document comprises the Group's activities and performances Report and the 'EDF at a glance' 2017 report

EPRI PWR primary water chemistry guidelines revision

International Nuclear Information System (INIS)

McElrath, Joel; Fruzzetti, Keith

2014-01-01

EPRI periodically updates the PWR Primary Water Chemistry Guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). The last revision of the PWR water chemistry guidelines identified an optimum primary water chemistry program based on then-current understanding of research and field information. This new revision provides further details with regard to primary water stress corrosion cracking (PWSCC), fuel integrity, and shutdown dose rates. A committee of industry experts, including utility specialists, nuclear steam supply system (NSSS) and fuel vendor representatives, Institute of Nuclear Power Operations (INPO) representatives, consultants, and EPRI staff collaborated in reviewing the available data on primary water chemistry, reactor water coolant system materials issues, fuel integrity and performance issues, and radiation dose rate issues. From the data, the committee updated the water chemistry guidelines that all PWR nuclear plants should adopt. The committee revised guidance with regard to optimization to reflect industry experience gained since the publication of Revision 6. Among the changes, the technical information regarding the impact of zinc injection on PWSCC initiation and dose rate reduction has been updated to reflect the current level of knowledge within the industry. Similarly, industry experience with elevated lithium concentrations with regard to fuel performance and radiation dose rates has been updated to reflect data collected to date. Recognizing that each nuclear plant owner has a unique set of design, operating, and corporate concerns, the guidelines committee has retained a method for plant-specific optimization. Revision 7 of the Pressurized Water Reactor Primary Water Chemistry Guidelines provides guidance for PWR primary systems of all manufacture and design. The guidelines continue to emphasize plant

Economy and the fuel market

International Nuclear Information System (INIS)

1994-01-01

The nuclear fuel manufacturing constitutes a considerable venture for the competitiveness of the nuclear power sector although it represents a relatively modest fraction (around 4%) of the nuclear kWh cost. The COGEMA group is participating through its branches in the control of the most part (32%) of the world manufacturing capacity of fuel for PWR. Amounting up to 242 operating installations this reactor type is the most widespread in the world. The paper discusses the costs, the fuel clients and the fuel suppliers. Data concerning the boiling water and fast neutron reactors, geographical localization of the PWR and VVER reactors all over the world, the PWR and fuel for PWR manufacturers are also presented

Utilization of thorium in PWR type reactors

International Nuclear Information System (INIS)

Correa, F.

1977-01-01

Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

Influence of boron reduction strategies on PWR accident management flexibility

International Nuclear Information System (INIS)

Papukchiev, Angel Aleksandrov; Liu, Yubo; Schaefer, Anselm

2007-01-01

In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. Design changes to reduce boron concentration in the reactor coolant are of general interest regarding three aspects - improved reactivity feedback properties, lower impact of boron dilution scenarios on PWR safety and eventually more flexible accident management procedures. In order to assess the potential advantages through the introduction of boron reduction strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 ppm and 805 ppm. For the assessment of the potential safety advantages of these cores a hypothetical beyond design basis accident has been simulated with the system code ATHLET. The analyses showed improved inherent safety and increased accident management flexibility of the low boron cores in comparison with the standard PWR. (author)

ROX PWR

International Nuclear Information System (INIS)

Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H.

1999-01-01

For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO 2 , ThO 2 and Er 2 O 3 , and a heterogeneous core with Zr-ROX and UO 2 assemblies. As a result, the additives UO 2 + Er 2 O 3 are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO 2 heterogeneous core, further reduction of power peaking seems necessary. (author)

Analysis of the performance of fuel cells PWR with a single enrichment and radial distribution of enrichments

International Nuclear Information System (INIS)

Vargas, S.; Gonzalez, J. A.; Alonso, G.; Del Valle, E.; Xolocostli M, J. V.

2008-01-01

One of the main challenges in the design of fuel assemblies is the efficient use of uranium achieving burnt homogeneous of the fuel rods as well as the burnt maximum possible of the same ones to the unload. In the case of the assemblies type PWR has been decided actually for fuel assemblies with a single radial enrichment. The present work has like effect to show the because of this decision, reason why a comparison of the neutronic performance of two fuel cells takes place with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The results shown in the present study of the behavior of the neutron flow as well as the power distribution through of assembly sustain the because of a single radial enrichment. (Author)

Study and Construction of the Metal Vessels for the Reactors of the EDF1 and EDF2 Sectors at Chinon; Etude et construction des caissons metalliques des reacteurs des tranches EDF1 et EDF2 de la centrale de Chinon; Izuchenie i konstruktsiya metallicheskikh korpusov reaktorov pervoj i vtoroj chasti programm ehlektrostantsij; Estudio y construccion de los recipientes metalicos de los reactores EDF1 y EDF2 de la central de Chinon

Energy Technology Data Exchange (ETDEWEB)

Lamiral, G.; Millot, R.; Passerieux, P. [Electricite de France, Clamart, Seine (France)

1963-10-15

The first two natural uranium-graphite-C0{sub 2} reactors at the Chinon station have metal vessels of thick manganese-molybdenum steel plate. The studies carried out on these vessels raised certain problems, particularly in connection with the design and dimensions of the port reinforcements. The reinforcements for the control-rod channels and fuel ports were studied on mock-ups and the results obtained were checked on the completed reactors during hydraulic tests. The type of construction initially used for the EDF1 vessel was relatively simple. The plates to be welded were locally preheated, and the vessel was not supposed to undergo more than one stress-relief heat treatment after completion of all the welding. Serious cracks developed, however, and it became necessary to alter the whole method of construction. In particular, the welding was now done after overall preheating and the vessel was subjected to multiple stress-relief treatments. This made it possible to fabricate the vessels for EDF1 and EDF2, but at the same time imposed certain limitations which considerably complicated work on the site. (author) [French] Les reacteurs a uranium naturel, graphite et gaz carbonique des deux premieres tranches de la Centrale de Chinon comportent des caissons metalliques realises a partir de toles de fortes epaisseurs, en acier au manganese-molybdene. Les etudes de ces paissons ont pose certains problemes, notamment en ce qui concerne les renforts d'ouvertures. Les renforts des passages des barres de controle et des orifices de chargement ont ete etudies sur maquette et les resultats obtenus ont ete controles sur les ouvrages termines lors des epreuves hydrauliques. Le mode de construction initialement utilise pour le caisson de la tranche EDF1 etait relativement simple; les toles a souder etaient prechauffees localement et le caisson ne devait subir qu'un seul traitement thermique de detente, apres execution de toutes les soudures. Une fissuration importante en cours

Advanced high conversion PWR: preliminary analysis

International Nuclear Information System (INIS)

Golfier, H.; Bellanger, V.; Bergeron, A.; Dolci, F.; Gastaldi, B.; Koberl, O.; Mignot, G.; Thevenot, C.

2007-01-01

In this paper, physical aspects of a HCPWR (High Conversion Light Water Reactor), which is an innovative PWR fuelled with mixed oxide and having a higher conversion ratio due to a lower moderation ratio. Moderation ratios lower than unity are considered which has led to low moderation PWR fuel assembly designs. The objectives of this parametric study are to define a feasibility area with regard to the following neutronic aspects: moderation ratio, Pu loading, reactor spectrum, irradiation time, and neutronic coefficients. Important thermohydraulic parameters are the pressure drop, the critical heat flux, the maximum temperature in the fuel rod and the pumping power. The thermohydraulic analysis shows that a range of moderation ratios from 0.8 to 1.2 is technically possible. A compromise between improved fuel utilization and research and development effort has been found for the moderation ration of about 1. The parametric study shows that there are 2 ranges of interest for the moderation ratio: -) moderation ratio between 0.8 and 1.2 with reduced fissile heights (> 3 m), hexagonal arrangement fuel assembly and square arrangement fuel assembly are possible; and -) moderation between 0.6 and 0.7 with a modification of the reactor operating conditions (reduction of the primary flow and of the thermal power), the fuel rods could be arranged inside a hexagonal fuel rod assembly. (A.C.)

Study of radiation effects on zircaloy 4 microstructure (Impact on susceptibility to fuel pellet-cladding interaction in PWR)

International Nuclear Information System (INIS)

Lefebvre, F.

1989-01-01

In PWR the fast neutron flux is an important parameter for fuel can aging by modification of zircaloy-4 microstructure: amorphisation and dissolution of intermetallic precipitates. These phenomena are both analysed and their influence on fuel-cladding interaction is discussed. Irradiations by 1 MeV electrons, Ar ions, Kr ions and fast neutrons are realized for comparison of damages with different defect creation kinetics. Amorphisation is explained as the crystal amorphous state transformation allowing precipitate dissolution by creation of a chemical potential gradient between matrix and amorphous phase. Progressive dissolution of precipitates produced by irradiation decrease the number of potential sites for stress corrosion cracking, improving rupture resistance of the alloy by fuel-cladding interaction [fr

At the heart of EDF's education plant

International Nuclear Information System (INIS)

Maillard, C.

2010-01-01

Facing a high and growing number of retirements and the present revival of nuclear energy, Electricite de France (EDF), the French state-owned electric utility, is struggling to find the competencies it needs, especially in domains such as reactor operators, mechanic technicians and engineers in the fields of fittings, boiler fabrication, maintenance, automation, etc. EDF is thus engaged in an important education and training program. The author reports from Bugey (France), where is located one of the main EDF production-engineering training center, and describes the variety of learning activities available in the center. An important effort is devoted to the training of foreign operators in order for EDF to better compete on the nuclear energy international market

The public utility contract between the French government and EDF; Le contrat de service public entre l'Etat et EDF

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-07-01

The public utility contract as foreseen by article 1 of the law from August 9, 2004 was signed on October 24, 2005. It represents the reference of EDF's (Electricite de France) commitments for its different activities of power producer-supplier and network manager. It precises also the commitments of the manager of the public energy transportation system (RTE EDF-Transport). It describes the level of commitment of EDF group over the 2005-2006-2007 era and precises the financial compensation modalities of its public utility missions and of its commitments as described in the contract. The public utility plan indicates the level of investments planned by EDF in the domain of power generation and networks. This document gathers: the talk given by D. de Villepin, Prime Minister, during the signature of the contract, the talk and presentation (slides) given by P. Gadonneix, head of EDF, the public utility contract between the government and EDF, and the public utility plan. (J.S.)

Comparison of DUPIC fuel composition heterogeneity control methods

International Nuclear Information System (INIS)

Choi, Hang Bok; Ko, Won Il

1999-08-01

A method to reduce the fuel composition heterogeneity effect on the core performance parameters has been studied for the DUPIC fuel which is made of spent pressurized water reactor (PWR) fuels by a dry refabrication process. This study focuses on the reactivity control method which uses either slightly enriched, depleted, or natural uranium to minimize the cost rise effect on the manufacturing of DUPIC fuel, when adjusting the excess reactivity control by slightly enriched and depleted uranium, reactivity control by natural uranium for high reactivity spent PWR fuels, and reactivity control by natural uranium for linear reactivity spent PWR fuels. The results of this study have shown that the reactivity control by slightly enriched and depleted uranium, all the spent PWR fuels can be utilized as the DUPIC fuel and the fraction of fresh uranium feed is 3.4% on an average. For the reactivity control by natural uranium, about 88% of spent PWR fuel can be utilized as the DUPIC fuel when the linear reactivity spent PWR fuels are used, and the amount of natural uranium feed needed to control the DUPIC fuel reactivity is negligible. (author). 13 refs., 16 tabs., 6 figs

Modelling of pellet-cladding interaction in PWR's

International Nuclear Information System (INIS)

Esteves, A.M.; Silva, A.T. e.

1992-01-01

The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyses the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. (author)

EDF. Group dynamics and activities. Competitive environment and strategic perspectives. Release - October 2017

International Nuclear Information System (INIS)

2017-10-01

After a synthesis, this report proposes a presentation of the EDF Group (general overview, activities, human resources, share-holding structure, stock market data). It gives an overview of the EDF Group dynamics and of its activities: environment analysis (world electric power production, power consumption in France, regulated and spot prices, turnover in France and per area and market segment), performance analysis, and competitive analysis (comparison with the main European energy companies). It analyses the different development axes and discusses main events regarding the consolidation of nuclear activities, investments in renewable energies, withdrawal from coal and fuel, diversification in energy services, and financial consolidation. Financial data are presented along with the main economic and financial indicators. Important statistical data are provided

PWR water chemistry controls: a perspective on industry initiatives and trends relative to operating experience and the EPRI PWR water chemistry guidelines

International Nuclear Information System (INIS)

Fruzzetti, K.; Choi, S.; Haas, C.; Pender, M.; Perkins, D.

2010-01-01

An effective PWR water chemistry control program must address the following goals: Minimize materials degradation (e.g., PWSCC, corrosion of fuel, corrosion damage of steam generator (SG) tubes); Maintain fuel integrity and good performance; Minimize corrosion product transport (e.g., transport and deposition on the fuel, transport into the SGs where it can foul tube surfaces and create crevice environments for the concentration of corrosive impurities); Minimize dose rates. Water chemistry control must be optimized to provide overall improvement considering the sometimes variant constraints of the goals listed above. New technologies are developed for continued mitigation of materials degradation, continued fuel integrity and good performance, continued reduction of corrosion product transport, and continued minimization of plant dose rates. The EPRI chemistry program, in coordination with other EPRI programs, strives to improve these areas through application of chemistry initiatives, focusing on these goals. This paper highlights the major initiatives and issues with respect to PWR primary and secondary system chemistry and outlines the recent, on-going, and proposed work to effectively address them. These initiatives are presented in light of recent operating experience, as derived from EPRI's PWR chemistry monitoring and assessment program, and EPRI's water chemistry guidelines. (author)

70 years of EDF archives and a lot more

International Nuclear Information System (INIS)

Holgado, S.

2016-01-01

Since 1968 'EDF Archives' has been in charge of collecting and preserving EDF's documents. EDF distinguishes 2 types of documents: documents (including data) that are still used by EDF teams in their daily work for knowledge transmission or legal aspects and earlier materials that are used for research works in history, sociology or economics. EDF being a public limited liability company (SA) with a public service mission, its archives are both private and public. The oldest documents date back in the 19. century as a legacy of the 2400 enterprises producing or transporting or dispatching electricity or gas that were nationalized in 1946 to form the EDF company. EDF's archives represent 100 linear kilometers of documents, 8 kilometers of which can be consulted by the public. In 2015 a club of the archivists from the nuclear industry was created, its aims were to manage common archives, to build a mapping of all corporate archives and to harmonize managing rules. (A.C.)

Comparison of DUPIC fuel composition heterogeneity control methods

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Ko, Won Il [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-08-01

A method to reduce the fuel composition heterogeneity effect on the core performance parameters has been studied for the DUPIC fuel which is made of spent pressurized water reactor (PWR) fuels by a dry refabrication process. This study focuses on the reactivity control method which uses either slightly enriched, depleted, or natural uranium to minimize the cost rise effect on the manufacturing of DUPIC fuel, when adjusting the excess reactivity of the spent PWR fuel. In order to reduce the variation of isotopic composition of the DUPIC fuel, the inter-assembly mixing operation was taken three times. Then, three options have been considered: reactivity control by slightly enriched and depleted uranium, reactivity control by natural uranium for high reactivity spent PWR fuels, and reactivity control by natural uranium for linear reactivity spent PWR fuels. The results of this study have shown that the reactivity of DUPIC fuel can be tightly controlled with the minimum amount of fresh uranium feed. For the reactivity control by slightly enriched and depleted uranium, all the spent PWR fuels can be utilized as the DUPIC fuel and the fraction of fresh uranium feed is 3.4% on an average. For the reactivity control by natural uranium, about 88% of spent PWR fuel can be utilized as the DUPIC fuel when the linear reactivity spent PWR fuels are used, and the amount of natural uranium feed needed to control the DUPIC fuel reactivity is negligible. 13 refs., 6 figs., 16 tabs. (Author)

EDF Group - Annual Report 2012

International Nuclear Information System (INIS)

2013-01-01

The EDF Group is one of the world's leading energy companies, active in all areas from generation to trading and network management. It has a sound business model, evenly balanced between regulated and deregulated activities. With its first-rate human resources, R and D capability, expertise in engineering and operating generation plants and networks, as well as its energy eco-efficiency offers, the Group delivers competitive solutions that help ensure sustainable economic development and climate protection. The EDF Group is the leader in the French and UK electricity markets and has solid positions in Italy and numerous other European countries, as well as industrial operations in Asia and the United States. Everywhere it operates, the Group is a model of quality public service for the energy sector. This document is EDF Group's annual report for the year 2012. It contains information about Group profile, governance, business, development strategy, sales and marketing, positions in Europe and international activities. The document is made of several reports: the Activity and Sustainable Development Report, the Financial Report, the 'EDF at a glance' report, and the Sustainable Development Indicators

Evaluation of burnup credit for accommodating PWR spent nuclear fuel in high-capacity cask designs

International Nuclear Information System (INIS)

Wagner, John C.

2003-01-01

This paper presents an evaluation of the amount of burnup credit needed for high-density casks to transport the current U.S. inventory of commercial spent nuclear fuel (SNF) assemblies. A prototypic 32-assembly cask and the current regulatory guidance were used as bases for this evaluation. By comparing actual pressurized-water-reactor (PWR) discharge data (i.e., fuel burnup and initial enrichment specifications for fuel assemblies discharged from U.S. PWRs) with actinide-only-based loading curves, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of SNF assemblies in high-capacity storage and transportation casks. The impact of varying selected calculational assumptions is also investigated, and considerable improvement in effectiveness is shown with the inclusion of the principal fission products (FPs) and minor actinides and the use of a bounding best-estimate approach for isotopic validation. Given sufficient data for validation, the most significant component that would improve accuracy, and subsequently enhance the utilization of burnup credit, is the inclusion of FPs. (author)

Studies of fuel loading pattern optimization for a typical pressurized water reactor (PWR) using improved pivot particle swarm method

International Nuclear Information System (INIS)

Liu, Shichang; Cai, Jiejin

2012-01-01

Highlights: ► The mathematical model of loading pattern problems for PWR has been established. ► IPPSO was integrated with ‘donjon’ and ‘dragon’ into fuel arrangement optimizing code. ► The novel method showed highly efficiency for the LP problems. ► The core effective multiplication factor increases by about 10% in simulation cases. ► The power peaking factor decreases by about 0.6% in simulation cases. -- Abstract: An in-core fuel reload design tool using the improved pivot particle swarm method was developed for the loading pattern optimization problems in a typical PWR, such as Daya Bay Nuclear Power Plant. The discrete, multi-objective improved pivot particle swarm optimization, was integrated with the in-core physics calculation code ‘donjon’ based on finite element method, and assemblies’ group constant calculation code ‘dragon’, composing the optimization code for fuel arrangement. The codes of both ‘donjon’ and ‘dragon’ were programmed by Institute of Nuclear Engineering of Polytechnique Montréal, Canada. This optimization code was aiming to maximize the core effective multiplication factor (Keff), while keeping the local power peaking factor (Ppf) lower than a predetermined value to maintain fuel integrity. At last, the code was applied to the first cycle loading of Daya Bay Nuclear Power Plant. The result showed that, compared with the reference loading pattern design, the core effective multiplication factor increased by 9.6%, while the power peaking factor decreased by 0.6%, meeting the safety requirement.

Modeling of PWR fuel at extended burnup; Estudo de modelos para o comportamento a altas queimas de varetas combustiveis de reatores a agua leve pressurizada

Energy Technology Data Exchange (ETDEWEB)

Dias, Raphael Mejias

2016-11-01

This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

Study of corium radial spreading between fuel rods in a PWR core

International Nuclear Information System (INIS)

Roche, S.; Gatt, J.M.

1996-01-01

In the framework of severe accident studies for PWR like Three Mile Island Unit 2 (TMI-2), the reactor core essentially constituted of fuel rods begins to heat and then to melt. During the early degradation phase, a melt (essentially UO2 and ZrO2) that constitutes the corium flows first along the rods, and after a blockage formation, may radially propagate towards the core periphery. A simplified model has been elaborated to study the corium freezing phenomena during its crossflow between the fuel rods. The corium spreads on an horizontal support made, of either a corium crust, or a grid assembly. The model solves numerically the interface energy balance equation at the solid-liquid corium interface and the monodimensional heat balance equation in transient process with convective terms and heat source (residual power). ''Zukauskas'' correlations are used to calculate heat transfer coefficients. The model can be integrated in severe accident codes like ICARE II (IPSN) describing the in-vessel degradation scenarios. (author). 5 refs, 10 figs

Validation of gadolinium burnout using PWR benchmark specification

Energy Technology Data Exchange (ETDEWEB)

Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

2014-07-01

Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

PWR fuel thermomechanics

International Nuclear Information System (INIS)

Traccucci, R.; Leclercq, J.

1986-01-01

Fuel thermo-mechanics means the studies of mechanical and thermal effects, and more generally, the studies of the behavior of the fuel assembly under stresses including thermal and mechanical loads, hydraulic effects and phenomena induced by materials irradiation. This paper describes the studies dealing with the fuel assembly behavior, first in normal operating conditions, and then in accidental conditions. 43 refs [fr

A model finite-element to analyse the mechanical behavior of a PWR fuel rod

International Nuclear Information System (INIS)

Galeao, A.C.N.R.; Tanajura, C.A.S.

1988-01-01

A model to analyse the mechanical behavior of a PWR fuel rod is presented. We drew our attention to the phenomenon of pellet-pellet and pellet-cladding contact by taking advantage of an elastic model which include the effects of thermal gradients, cladding internal and external pressures, swelling and initial relocation. The problem of contact gives rise ro a variational formulation which employs Lagrangian multipliers. An iterative scheme is constructed and the finite element method is applied to obtain the numerical solution. Some results and comments are presented to examine the performance of the model. (author) [pt

Inspection systems for valves monitoring at EDF

International Nuclear Information System (INIS)

Germain, J.L.; Granal, L.; Provost, D.; Touillez, M.

1997-01-01

Electricite de France (EDF) makes increasing use of valve inspection systems to guarantee safety in its pressurized water reactor plants, improve plant availability and facilitate condition-based maintenance. A portable system known as SAMIR has been developed for inspection of motor-operated valves, and is now used on EDF's 900-MW sites. For its 1300-MW units, EDF has chosen a more complete system which enables measuring thrust on the valve stem during a maneuver, using a sensor mounted on the yoke. To detect internal vale leaks, an on-site assessment has demonstrated the economic benefits of acoustic emission techniques. EDF has equipped its sites with analog leak detection systems which may soon be replaced by a digital model now being developed. (authors)

ROX PWR

Energy Technology Data Exchange (ETDEWEB)

Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1999-12-01

For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO{sub 2}, ThO{sub 2} and Er{sub 2}O{sub 3}, and a heterogeneous core with Zr-ROX and UO{sub 2} assemblies. As a result, the additives UO{sub 2}+ Er{sub 2}O{sub 3} are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO{sub 2} heterogeneous core, further reduction of power peaking seems necessary. (author)

Optimization of reload core design for PWR

International Nuclear Information System (INIS)

Shen Wei; Xie Zhongsheng; Yin Banghua

1995-01-01

A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

Study on Reactor Physics Characteristic of the PWR Core Using UO2

International Nuclear Information System (INIS)

Tukiran Surbakti

2009-01-01

Study on reactor physics characteristic of the PWR core using UO 2 fuel it is necessary to be done to know the characteristic of geometry, condition and configuration of pin cell in the fuel assembly Because the geometry, configuration and condition of the pin cell in fuel core determine the loading strategy of in-core fuel management Calculation of k e ff is a part of the neutronic core parameter calculation to know the reactor physics characteristic. Generally, core calculation is done using computer code starts from modelling one unit fuel lattice cell, fuel assembly, reflector, irradiation facility and until core reactor. In this research, the modelling of pin cell and fuel assembly of the PWR 17 ×17 is done homogeneously. Calculation of the k-eff is done with variation of the fuel volume fraction, fuel pin diameter, fuel enrichment. The calculation is using by NITAWL and CENTRM, and then the results will be compared to KENOVI code. The result showed that the value of k e ff for pin cell and fuel assembly PWR 17 ×17 is not different significantly with homogenous and heterogenous models. The results for fuel volume fraction of 0.5; rod pitch 1.26 cm and fuel pin diameter of 9.6 mm is critical with burn up of 35,0 GWd/t. The modeling and calculation method accurately is needed to calculation the core physic parameter, but sometimes, it is needed along time to calculate one model. (author)

Improved identification to prevent transposition during operation of 900 MWe PWR reactors

International Nuclear Information System (INIS)

Leckner, J.M.; Dien, Y.; Cernes, A.

1986-04-01

Detailed human factors analysis of 900 MWe PWR control room identification systems was carried out by the Nuclear and Fossil Generation Division of Electricite de France (EDF) consequent to a series of incidents where personnel confused one plant unit, room or piece of equipment for another. Preliminary analysis uncovered coding inadequacies and suggested possible remedies. This data was used to prepare specifications for identification redesign at a pilot plant on which detailed investigations could be carried out. Recommended solutions were submitted to pilot plant operators and their opinion sollicited. Operator recommendations will be tried out on the pilot plant and adopted on a grid-wide basis if trials prove satisfactory

The study on radioactivity reduction of spent PWR cladding hull

International Nuclear Information System (INIS)

Jung, I. H.; Kim, J. H.; Park, C. J.; Jung, Y. H.; Song, K. C.; Lee, J. W.; Park, J. J.; Yang, M. S.

2003-01-01

Hull arising from the spent PWR fuel elements is classified as a high-level radioactive waste. This report describes the radio-chemical characteristics of the hull-from PWR spent fuel of 32,000MWd/tU burn-up and 15 years cooling, discharged from Gori Unit I cycled 4-7-by examination and literature survey. On the basis of the results, a method of degradation to middle and low-level radioactive waste was proposed by dry process such as laser or plasma technique with removing the nuclides deposited on the surface of the hull

DIADEME: A computer code to assess in operation defective fuel characteristics and primary circuit contamination

Energy Technology Data Exchange (ETDEWEB)

Genin, J.B. [DEN/DEC/S3C, CEA Cadarache, 13 - Saint-Paul-lez-Durance (France); Harrer, A. [EdF/SEPTEN, 69 - Villeurbanne (France); Musante, Y. [FRAMATOME-ANP, 69 - Lyon (France)

2002-07-01

DIADEME is a computer code developed within the framework of R and D cooperation between the French Atomic Energy Commission (CEA), Electricite de France (EdF) and FRAMATOME-ANP. Its aim is to assess in operation defective fuel characteristics and primary circuit contamination for actinides and long half-life fission products involved in health physics problems as well as in waste and decommissioning studies. DIADEME has been developed and qualified for the EDF nuclear power plants. For many years, both theoretical and experimental studies have been carried out at the CEA on the release of fission products and actinides out of defective fuel rods in operation, their migration and deposition in PWR primary circuits. These studies have allowed defect characteristic diagnosis methods to be developed, based on radiochemical measurements of the primary coolant. These methods are generally used along with gamma spectrometry measurements on primary water sampling. In order to be completely efficient, these methods can also be used in connection with an on-line primary water gamma spectrometry device. This permits to obtain the most comprehensive data on fission product activity evolutions at steady state and during operation transients, and allows the on-line characterization of the defective fuel assemblies. For long half-life fission products and for actinides, DIADEME is also able to assess the activities of soluble and insoluble forms in the primary water and in the chemical and voluminal control system (CVCS) filters and resins, as well as those activities deposited on primary circuit surfaces. (author)

DIADEME: A computer code to assess in operation defective fuel characteristics and primary circuit contamination

International Nuclear Information System (INIS)

Genin, J.B.; Harrer, A.; Musante, Y.

2002-01-01

DIADEME is a computer code developed within the framework of R and D cooperation between the French Atomic Energy Commission (CEA), Electricite de France (EdF) and FRAMATOME-ANP. Its aim is to assess in operation defective fuel characteristics and primary circuit contamination for actinides and long half-life fission products involved in health physics problems as well as in waste and decommissioning studies. DIADEME has been developed and qualified for the EDF nuclear power plants. For many years, both theoretical and experimental studies have been carried out at the CEA on the release of fission products and actinides out of defective fuel rods in operation, their migration and deposition in PWR primary circuits. These studies have allowed defect characteristic diagnosis methods to be developed, based on radiochemical measurements of the primary coolant. These methods are generally used along with gamma spectrometry measurements on primary water sampling. In order to be completely efficient, these methods can also be used in connection with an on-line primary water gamma spectrometry device. This permits to obtain the most comprehensive data on fission product activity evolutions at steady state and during operation transients, and allows the on-line characterization of the defective fuel assemblies. For long half-life fission products and for actinides, DIADEME is also able to assess the activities of soluble and insoluble forms in the primary water and in the chemical and voluminal control system (CVCS) filters and resins, as well as those activities deposited on primary circuit surfaces. (author)

Study on the behavior of waterside corroded PWR fuel rods under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Sasajima, Hideo

1989-06-01

One of the highlighted problems from the fuel reliability point of view is a waterside corrosion of fuel cladding which becomes more significant at extended burnup stages. To date, at highly burned fuel, waterside corrosion was recognized as important because cladding oxidation increased with increasing burn-up. In experiments, as the basic research for the study of high burn-up fuel, the test fuel rods were prepressurized to ranges from 3.47 to 3.55 MPa, oxidized artificially to both 10 and 20 μm in thickness. Regarding fabricated oxide thickness of 10 μm, it is corresponded to be transition point from cubic law to linear law as a function of burn-up. Pulse irradiation experiments by NSRR were carried out to study the behavior of waterside corroded PWR type fuels under RIA conditions. Obtained results are: (1) The failure threshold of tested fuels was 110 cal/g·fuel (0.46 KJ/g·fuel) in enthalpy. This showed that the failure threshold of tested fuels was same as that of the past NSRR experimental data. (2) The failure mechanisms of the tested fuel rods was cladding rupture induced by ballooning. No differences in failure mechanisms existed between the past NSRR prepressurized standard fuel and the tested fuels. (3) Cracks were existed without propagating into cladding matrix, so that it was judged that these were not initiation of failure. (4) Whithin this experimental condition, reduction of cladding thickness being attributed to the increase of oxidation did not failure threshold. (author)

Immersed multiple device for the control of the irradiated PWR fuel pins in the reloadable loop in the OSIRIS pond

International Nuclear Information System (INIS)

Farny, G.

1983-01-01

With respect to the dynamics of the degradation of the PWR fuel in transient, normal and abnormal regions, a new multi-device immersed in the cooling pond of the OSIRIS reactor, is studied. The multiple device is subjected to three examinations: (1) visual studying and video-recording of the appearance of the fuel pins, (2) metrology of the pins, (3) investigation of the induced Foucault currents in the fuel cans. Attention is chiefly paid to the last point; the other ones - being closely related - are only touched on whenever needed. It is concluded that quality control of the fuel pins is possible by means of Foucault currents without applying mechanical constraints and without interfering with the cooling rate. (Auth.)

Seismic isolation of nuclear power plants - EDF's philosophy

International Nuclear Information System (INIS)

Coladant, C.

1989-01-01

The elastomer bearing pads used since 1963 as supports for prestressed concrete pressure vessels (PCPVs) was quickly chosen by Electricite de France (ED) to improve the capability of nuclear power plants (NPPs) to withstand strong earthquakes and to reduce the seismic loads on structures and equipment. The standardized units for 900 and 1,300 MW(e) pressurized water reactor (PWR) plants have moderate seismic design loads of 0.2 and 0.15 g, respectively. These design loads were exceeded by the site dependent spectra of Cruas (France) and Koeberg (South Africa). To keep the plant design unchanged and to take the advantages of standardization, these units were put on laminated bearings with or without sliding plates. For the future French 1,500 MW(e) fast breeder reactors (FBRs), which are more sensitive to seismic loads, the base isolation is considered by EDF at the beginning of the design, even for low ground motions of 0.1 g. The buildings are placed on laminated bearings while the reactor block is supported by springs and dampers. The isolated plant has identical costs as a conventional design such as SPX1 at Creys-Malville

EDF group - annual report 2003

International Nuclear Information System (INIS)

2004-01-01

This document contains the magazine, the financial statements and the sustainable development report of Electricite de France (EdF) group for 2003: 1 - the magazine (chairman's statement, group profile, vision and strategy); 2 - the consolidated financial statements for the period ended 31 December 2003 (statutory auditors' report on the consolidated financial statements, EDF's summary annual financial statements); 3 - sustainable development report (transparency and dialogue, responsibility, commitment, partnerships for progress). (J.S.)

EDF Group - 2010 Reference Document

International Nuclear Information System (INIS)

2011-04-01

Beside the accounts of EDF for 2008 and 2009, this voluminous document presents persons in charge, legal account auditors, and how risks are managed within the company. It gives an overview of EDF activities, of its organization, of its assets. It presents and discusses its financial situation and results, indicates the main contracts, and proposes other documents concerning the company. Many documents and reports are provided in appendix

Integrating availability and maintenance objectives in plant design. EDF approach

International Nuclear Information System (INIS)

Degrave, Claude; Martin-Onraet, Michel

1995-01-01

Energy self sufficiency is a major strategic necessity for France. Regarding the fossil fuels power, competitiveness of nuclear energy is a key goal for Electricite de France. Accordingly, for future nuclear power plants to remain competitive, it is necessary to maintain the kWh production costs of the future units at a level close to those of the latest units under construction (N4 series), while raising the safety level. EDF therefore decided to implement an analytical and systematic process for study of the new projects to optimize the design by integration of the maintenance (durations, costs), availability and radiation exposure goals from the related operating experience. This approach, CIDEM (French acronym for Design Integrating Availability, operating Experience and Maintenance) aims at a single goal: to minimize the kWh production cost incorporating investment, operation and fuel costs, allowing for the operating experience from French and foreign units. The implementation of the CIDEM process constitutes for EDF a new approach to the study of the new Nuclear Power Plant projects. The competitivity of nuclear energy greatly depends on the success of such an approach. The studies conducted in the availability field have already highlighted a number of critical points and have made it possible to define the corresponding goal allocations and to establish a first series of structuring specifications for the project. (J.P.N.)

Contribution to the study of the conversion PWR type reactors to the thorium cycle

International Nuclear Information System (INIS)

Martins Filho, J.R.

1980-01-01

The use of the thorium cycle in PWR reactors is discussed. The fuel has been calculated in the equilibrium condition for a economic comparison with the uranium cycle (in the same condition). First of all, a code named EQUILIBRIO has been developed for the fuel equilibrium calculation. The results gotten by this code, were introduced in the LEOPARD code for the fuel depletion calculation (in the equilibrium cycle). Same important physics details of fuel depletion are studied, for instance: the neutron balance, power sharing, fuel burnup, etc. The calculations have been done taking as reference the Angra-1 PWR reactor. (Author) [pt

Parameterized representation of macroscopic cross section for PWR reactor

International Nuclear Information System (INIS)

Fiel, João Cláudio Batista; Carvalho da Silva, Fernando; Senra Martinez, Aquilino; Leal, Luiz C.

2015-01-01

Highlights: • This work describes a parameterized representation of the homogenized macroscopic cross section for PWR reactor. • Parameterization enables a quick determination of problem-dependent cross-sections to be used in few group calculations. • This work allows generating group cross-section data to perform PWR core calculations without computer code calculations. - Abstract: The purpose of this work is to describe, by means of Chebyshev polynomials, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 92 U enrichment. The cross-section data analyzed are fission, scattering, total, transport, absorption and capture. The parameterization enables a quick and easy determination of problem-dependent cross-sections to be used in few group calculations. The methodology presented in this paper will allow generation of group cross-section data from stored polynomials to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by the proposed methodology when compared with results from the SCALE code calculations show very good agreement

Analyses of PWR spent fuel composition using SCALE and SWAT code systems to find correction factors for criticality safety applications adopting burnup credit

Energy Technology Data Exchange (ETDEWEB)

Shin, Hee Sung; Suyama, Kenya; Mochizuki, Hiroki; Okuno, Hiroshi; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-01-01

The isotopic composition calculations were performed for 26 spent fuel samples from the Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using the SAS2H module of the SCALE4.4 code system with 27, 44 and 238 group cross-section libraries and the SWAT code system with the 107 group cross-section library. For the analyses of samples from the Obrigheim PWR reactor, geometrical models were constructed for each of SCALE4.4/SAS2H and SWAT. For the analyses of samples from 7 PWR reactors, the geometrical model already adopted in the SCALE/SAS2H was directly converted to the model of SWAT. The four kinds of calculation results were compared with the measured data. For convenience, the ratio of the measured to calculated values was used as a parameter. When the ratio is less than unity, the calculation overestimates the measurement, and the ratio becomes closer to unity, they have a better agreement. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed, however: (1) Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from the Obrigheim reactor (10 samples were deselected because their burnups were measured with Cs-137 non-destructive method, less reliable than Nd-148 method the rest 16 samples were measured with); (2) Larger than unity ratios were found for Am-241 and Cm-242 for both the 16 and 55 samples; (3) Larger than unity ratios were found for Sm-149 for the 55 samples; (4) SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor

EDF Energies Nouvelles. Consolidated financial statements at 30 June 2009

International Nuclear Information System (INIS)

2010-01-01

EDF Energies Nouvelles is a world leader in renewable energy electricity. The company develops, builds and operates clean energy power plants both for its own account and for third parties. Historically, EDF Energies Nouvelles primarily developed its business in two geographical areas, Europe and North America (U.S., Canada and Mexico). EDF Energies Nouvelles is a subsidiary of EDF, helping the Group to achieve its renewable energy goals. The EDF Group generates low-carbon electricity around the world and actively participates in the energy transition. EDF Energies Nouvelles prioritizes development of wind and photovoltaic solar capacity. As an integrated operator with global reach, EDF Energies Nouvelles covers the entire renewable energy chain, from development to operation and maintenance, and manages all project phases in-house. This document is EDF Energies Nouvelles's registration document for the year 2009. It contains information about Group profile, governance, business, investments, property, plant and equipment, management, financial position, employees, shareholders, etc. The document includes the group's consolidated financial statements at 31 December 2008

Report of Post Irradiation Examination for Dry Process Fuel

International Nuclear Information System (INIS)

Par, Jang Jin; Jung, I. H.; Kang, K. H.; Moon, J. S.; Lee, C. R.; Ryu, H. J.; Song, K. C.; Yang, M. S.; Yoo, B. O.; Jung, Y. H.; Choo, Y. S.

2006-08-01

The spent PWR fuel typically contains 0.9 wt.% of fissile uranium and 0.6 wt.% of fissile plutonium, which exceeds the natural uranium fissile content of 0.711 wt.%. The neutron economy of a CANDU reactor is sufficient to utilize the DUPIC fuel, even though the neutron-absorbing fission products contained in the spent PWR fuel were remained in the DUPIC fuel. The DUPIC fuel cycle offers advantages to the countries operating both the PWR and CANDU reactors, such as saving the natural uranium, reducing the spent fuel in both PWR and CANDU, and acquiring the extra energy by reuse of the PWR spent fuel. This report contains the results of post-irradiation examination of the DUPIC fuel irradiated four times at HANARO from May 2000 to August 2006 present except the first irradiation test of simulated DUPIC fuel at HANARO on August 1999

Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR); Implementacion en software libre del simulador universitario de nucleoelectrica tipo PWR (SU-PWR)

Energy Technology Data Exchange (ETDEWEB)

Valle H, J.; Hidago H, F.; Morales S, J.B. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: julfi_jg@yahoo.com.mx

2007-07-01

Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

Basic information about development and construction of a PWR

International Nuclear Information System (INIS)

Meyer, P.J.

1977-01-01

1.0) Plant layout of a PWR; 2.0) principle design of a PWR and the reactor coolant system; 3.0) reactor auxiliary and ancillary systems; 3.1) volume control system; 3.2) boric acid control and chemical feeding system; 3.3) coolant purification and degassing system; 3.4) coolant storage and treatment system; 3.5) nuclear component cooling system; 3.6) liquid waste processing system; 3.7) gaseous waste processing system; 4.0) residual heat removal system; 5.0) emergency feedwater system; 6.0) containment design; 7.0) fuel handling, storage and transport system in a PWR. (orig.) [de

Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-04-01

This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

Hearing of Mr. Francois Roussely, President of EdF

International Nuclear Information System (INIS)

Roussely, F.

2005-01-01

This document is the proceedings of the hearing of F. Roussely, President of Electricite de France (EdF), at the commission of economic affairs of the French house of commons, about the advisability of the construction of the EPR (European pressurized reactor) demonstration plant and about its possible financing by EdF. In a first part, F. Roussely recalls the European context of deregulation of energy markets and its impact of the French electric power industry (opening of the French market, industrial and social actions of EdF, need of a new generation of nuclear reactor, preservation of EdF's energy mix, warranty of public utility, un-bundling between energy trade and distribution, EdF's turnover and profitability, EdF's foreign daughter companies). In a second part, F. Roussely answers a series of questions asked by the different members of the commission concerning the different points presented in the first part. (J.S.)

A study on the direct use of spent PWR fuel in CANDU -A study on the radioactive waste management for DUPIC fuel cycle-

International Nuclear Information System (INIS)

Park, Hyun Soo; Jun, Kwan Sik; Nah, Jung Won; Park, Jang Jin; Kim, Jong Hoh; Cho, Yung Hyun; Baek, Seung Woo; Shin, Jin Myung; Yang, Seung Yung

1994-07-01

The immobilization materials for radioactive wastes resulting from the DUPIC fuel manufacturing process were selected and their characteristics were evaluated. To predict the trapping behavior of the Ruthenium, a semi-volatile nuclide, its volatility was measured and thermogravimetric analysis were performed with simulated fuel. New Ruthenium trapping material was developed which is deposited on ceramic honey-comb monolith of cordierite. The base glass was manufactured with fly ash added to the borosilicate glass. The composition of the scrap waste was calculated based on the PWR spent fuel which has initial 235 U content of 3.5%, burnup of 35,000 MWD/MTU and cooling time of 10 years. Simulated waste glass was manufactured, and its chemical durability was evaluated by soxhlet leach test. Radioactivity of non-oxidized cladding material were measured. The preliminary design criteria were prepared for off-gas treatment system in IMEF. 31 figs, 42 tabs, 51 refs. (Author)

Experimental studies of resistance fretting-wear of fuel rods for VVER-1000 and TVS-KVADRAT fuel assemblies

International Nuclear Information System (INIS)

Makarov, V.; Afanasiev, A.; Egorov, Yu.; Matvienko, I.

2015-01-01

The paper covers the results of the studies performed to justify the wear resistance of fuel rods in contact with the spacer grids of TVS VVER-1000 fuel assembly and TVS-KVADRAT square fuel assembly of Russian design for PWR-900 reactor. The presented results of three testing stages comprise: Testing of mockup fuel rods of VVER TVS fuel assembly for fretting wear under the conditions of the water chemistry of VVER reactor; Testing models of different design embodiments of the fuel rods for VVER TVS fuel assembly for fretting wear in still cold water; Testing mockup fuel rods of TVS-KVADRAT square fuel assembly for PWR reactor for frettingwear under the conditions of PWR water chemistry. The effect of structural and operational factors was determined (amplitudes, fuel rod vibration frequencies, values of cladding-to-spacer grid cell gap for the depth of fuel rod cladding wear etc.), an assessment was made of the threshold values of fuel rod vibration parameters, which, if not exceeded, provide the absence of the fuel rod cladding fretting wear in the fuel rod-to spacer grid contact area. Key words: fretting wear, fuel rod, spacer grid, VVER, PWR (author)

Contribution to numerical and mechanical modelling of pellet-cladding interaction in nuclear reactor fuel rod

International Nuclear Information System (INIS)

Retel, V.

2002-12-01

Pressurised water reactor fuel rods (PWR) are the place of nuclear fission, resulting in unstable and radioactive elements. Today, the mechanical loading on the cladding is harder and harder and is partly due to the fuel pellet movement. Then, the mechanical behaviour of the cladding needs to be simulated with models allowing to assess realistic stress and strain fields for all the running conditions. Besides, the mechanical treatment of the fuel pellet needs to be improved. The study is part of a global way of improving the treatment of pellet-cladding interaction (PCI) in the 1D finite elements EDF code named CYRANO3. Non-axisymmetrical multidirectional effects have to be accounted for in a context of unidirectional axisymmetrical finite elements. The aim of this work is double. Firstly a model simulating the effect of stress concentration on the cladding, due to the opening of the radial cracks of fuel, had been added in the code. Then, the fragmented state of fuel material has been taken into account in the thermomechanical calculation, through a model which led the strain and stress relaxation in the pellet due to the fragmentation, be simulated. This model has been implemented in the code for two types of fuel behaviour: elastic and viscoplastic. (author)

Analysis of the loss of pool cooling accident in a PWR spent fuel pool with MAAP5

International Nuclear Information System (INIS)

Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

2014-01-01

Highlights: • A PWR spent fuel pool was modeled by using MAAP5. • Loss of pool cooling severe accident scenarios were studied. • Loss of pool cooling accidents with two mitigation measures were analyzed. - Abstract: The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents the analysis of loss of pool cooling accident scenarios and the discussion of mitigation measures for the SFP at a pressurized water reactor (PWR) NPP with the MAAP5 code. Analysis of uncompensated loss of water due to the loss of pool cooling with different initial pool water levels of 12.2 m (designated as a reference case) and 10.7 m have been performed based on a MAAP5 input model. Scenarios of the accident such as overheating of uncovered fuel assemblies, oxidation of claddings and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission products were predicted with the assumption that mitigation measures were unavailable. The results covered a broad spectrum of severe accident evaluations in the SFP. Furthermore, as important mitigation measures, the effects of recovering the SFP cooling system and makeup water in SFP on the accident progressions have also been investigated respectively based on the events of pool water boiling and spent fuels uncovery. Based upon the reference case, three cases with the recovery of SFP cooling system and three other cases with makeup water in SFP have been studied. The results showed that, severe accident might happen if SFP cooling system was not restored timely before the spent fuels started to become uncovered; spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate larger than the average evaporation rate defined as the division of pool water mass above the

EDF and GDF diversification; La diversification d'EDF et de GDF

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-11-08

In the last years Electricite de France (EDF) and Gaz de France (GDF) developed intervening strategies in a number of sectors opened to competition, the so-called strategies of diversification or complementary activities. These new strategies have multiple objectives as, for instance, prospective monopole evolution, increasing exigence of the consumers' demand, or else competence valuing. Ever since 1991, these strategic orientations developed with the support of public powers that worked out first measures of organization and monitoring. Although the frame was loose and rather tolerant, boosting EDF and GDF diversification was perceived by the concerned professionals as an aggression directed against them. Following a strong mobilization of professional organizations and a number of elected persons, in mid 1993, the minister of industry was impelled to tackle personally with the matter what resulted in an accurate system of framing. The GDF diversification is essentially characterized by a 'thermal pole' corresponding to the activities of heating operation. GDF controls, around the company Danto Rogeat and the old branches of Shell group, about 5% of a market of heating operation in France, i.e., represents the third place in a rather monopolized sector where the two main operators represent almost 75%. The EDF diversification is but more diffuse. Three main fields emerge distinctly: the waste processing, public lighting, and engineering. EDF is also present on more restraint markets as cartography, tele-surveillance, cable TV, aquaculture. The implementation of the system is presented where prerogatives of the Council of State on one hand and the Council of Competition on the other hand are stipulated. The disposition frame system is constituted of two measures stipulating the application of commendation from the two councils and provisions ensuring the transparency, respectively.

EDF feedback on recent EPRI SGOG SG chemical cleanings applications for TSP blockage reduction and heat transfer recover

International Nuclear Information System (INIS)

Dijoux, M.; De Bouvier, O.; Mercier, S.; Pages, D.; Bretelle, J.-L.; Leclercq, P.; Mermillod, A.

2010-01-01

Between 2007 and 2008, six Steam Generators Chemical Cleanings (SGCC) with the inhibitor free high temperature process were applied on EDF PWR units. The main goal was to reduce the excessive Tube Support Plate blockages observed on several units of the EDF fleet and the consequences on wide range levels and the risk of tube cracks. The heat transfer recovery was the second objective. Despite the correct results obtained, the corrosion impact of the high temperature process on internal metallic surfaces, higher than expected, and the environmental issues led EDF to move to a new cleaning process. The low temperature process developed by EPRI SGOG and applied for many years was selected for the same purpose. Some qualification laboratory tests were performed by Dominion Engineering Inc (DEI) to demonstrate the innocuousness an the efficiency of the process to achieve these goals. The EPRI SGOG process was then applied seven times by Westinghouse on the EDF units Cruas 3, Cruas 2, Belleville 1, Cattenom 1, Cattenom 3, Chinon B3 and Cattenom 4 between 2008 and 2010. All these units operate from the initial start at low AVT pH 25 o C (9,2) in the secondary circuit. Due to copper presence in the deposits to remove, the cleaning sequence 'Copper - Iron - Copper steps' was performed each time. After a short description of the process, including the specific adaptation in France, lessons learned are reported in this paper in the following areas: process monitoring, corrosion, efficiency, liquid and gaseous wastes, chemical pollution during start-up. Based on the 3 first applications in 2008, some modifications of the process were implemented, particularly for the copper step. For the units cleaned, 1100 to 4500 kg of deposits per SG have been removed, including TS sludge lancing. The reduction of TSP blockages was satisfying. The effect on steam pressure improvement and the wide range level is then discussed. The paper concludes on EDF perspectives for soft

EDF Group - Annual Report 2014. The people who power tomorrow

International Nuclear Information System (INIS)

2015-01-01

The EDF Group is the world's leading electricity company and very well established in Europe. Its business covers all electricity-related activities, from generation to networks and commerce. It is an important player in energy trading through EDF trading. This document is EDF Group's annual report for the year 2014. It contains information about Group profile, governance, business, development strategy, sales and marketing, positions in Europe and international activities. The document is made of several reports: the Activity and Sustainable Development Report, the 'EDF at a glance' report, and the EDF Group Performance sheet

ABB high burnup fuel

International Nuclear Information System (INIS)

Andersson, S.; Helmersson, S.; Nilsson, S.; Jourdain, P.; Karlsson, L.; Limback, M.; Garde, A.M.

1999-01-01

Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both PWR and BWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter proven to meet the 6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10 x 10 fuel, where ABB is the only vendor to date with batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of PWR and BWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its utility customers. This paper provides an overview of recent fuel performance and reliability experience at ABB. Selected development and validation activities for PWR and BWR fuel are presented, for which the ABB test facilities in Windsor (TF-2 loop, mechanical test laboratory) and Vaesteras (FRIGG, BURE) are essential. (authors)

EDF - The Inspector General's report on Nuclear Safety and Radiation Protection

International Nuclear Information System (INIS)

Tandonnet, Jean

2012-01-01

This report is destined for the President of EDF with the purpose of informing him of the judgement the Inspector General has about the nuclear safety and radiation protection within the EDF Group. It is also intended for all who, in any way, play a role at EDF in improving nuclear safety. It may also prove useful to those outside the company who work in the nuclear field or indeed any branch of industry where there are risks to be managed and who wish to participate in the debate. It also contributes to the pool of information shared with the other nuclear operators under the auspices of WANO. The continuous improvement of nuclear safety results is a fundamental commitment of the EDF Group. This is notably reflected in a policy of transparency, whether in France, Britain, America or China. This document constitutes an example of this, offering an independent view of the operational situation. It is based on facts and findings derived in EDF's contacts with field staff and meetings with the decision makers, managers, medical personnel and players on both sides of this industry as well as outside stakeholders, especially contractor companies. The resulting snapshot of the situation in the field is particularly instructive. This report therefore concentrates more on problems and weaknesses rather than strengths and progress, except as concerns happenings outside France, where the focus is more on good practices than difficulties. Contents: 1 - The view from the Inspector General of EDF about the nuclear safety in the EDF group; 2 - A panorama of 2011: The management, The state of the installations, Maintenance of the French nuclear power plants, Relations with the French nuclear safety authority (ASN), The EPR projects, The players in the fields of health and safety, Radiation protection, Fire prevention and fire fighting, Increasing nuclear power plant security, Decommissioning; 3 - Nuclear safety management; 4 - The Fukushima Daiichi accident; 5 - Training to

The human factors approach at EDF

International Nuclear Information System (INIS)

Colas, A.

2004-01-01

At the dawn of the 21st century, French electricity utility EDF is facing a number of major changes, in particular the liberalisation of European energy markets and the restructuring needed to cope with this development. EDF's approach to human factors (HF) aspects is also undergoing major changes, since people obviously play a predominant role in any organisational structure. (author)

The DUPIC alternative for backend fuel cycle

International Nuclear Information System (INIS)

Lee, J.S.; Yang, M.S.; Park, H.S.; Boczar, P.; Sullivan, J.; Gadsby, R.D.

1997-01-01

The DUPIC fuel cycle was conceived as an alternative to the conventional fuel cycle backed options, with a view to multiple benefits expectable from burning spent PWR fuel again in CANDU reactors. It is based on the basic idea that the bulk of spent PWR fuel can be directly refabricated into a reusable fuel for CANDU of which high efficiency in neutron utilization would exhaustively burn the fissile remnants in the spent PWR fuel to a level below that of natural uranium. Such ''burn again'' strategy of the DUPIC fuel cycle implies that the spent PWR fuel will become CANDU fuel of higher burnup with relevant benefits such as spent PWR fuel disposition, saving of natural uranium fuel, etc. A salient feature of the DUPIC fuel cycle is neither the fissile content nor the bulk radioactivity is separated from the DUPIC mass flow which must be contained and shielded all along the cycle. This feature can be considered as a factor of proliferation resistance by deterrence against access to sensitive materials. It means also the requirement for remote systems technologies for DUPIC fuel operation. The conflicting aspects between better safeguardability and harder engineering problems of the radioactive fuel operation may be the important reason why the decades' old concept, since INFCE, of ''hot'' fuel cycle has not been pursued with much progress. In this context, the DUPIC fuel cycle could be a live example for development of proliferation resistant fuel cycle. As the DUPIC fuel cycle looks for synergism of fuel linkage from PWR to CANDU (or in broader sense LWR to HWR), Korea occupies a best position for DUPIC exercise with her unique strategy of reactor mix of both reactor types. But the DUPIC benefits can be extended to global bonus, expectable from successful development of the technology. (author)

Maintenance and repair tools of the 900 MW PWR units in France

International Nuclear Information System (INIS)

Blary, F.; Castaing, C.; Hassig, J.

1985-01-01

EDF and FRAMATOME have developed special tools and, repair and maintenance procedures to decrease the unavailability of the PWR units and the dosimetry of the personnel of maintenance. This report presents the solution adopted to face the problem of replacement of the internal upper structure pins of 21 units: 1) study and realisation of important industrial means; 2) development of a strategy: the pin replacement is separated from the guide tube replacement by the creation of a pin replacement shop, and, as often as it is possible, the ''AQUARIUM'' method allows to replace the guide-tubes with the aid of a mobile pool called AQUARIUM outside the critical way of shutdown [fr

EDF group. Annual report 2001

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-05-01

This document is the English version of the 2001 annual report of Electricite de France (EdF) Group, the French electric utility. It comprises 4 parts: introduction (statement of the chairman and chief executive officer, corporate governance, group key figures, sustainable growth indicators - parent company, energy for a sustainable future, EdF group worldwide); dynamics and balanced growth (financial results, EdF's strategy in building a competitive global group: consolidating the European network, moving forward in energy-related services, responding to increasing energy demand in emerging countries); sustainable solutions for all (empowering the customer: competitive solutions for industrial customers, anticipating the needs of residential customers and SMEs, environmental solutions to enhance urban life, upgrading the network and providing access to energy; a sound, sustainable and secure energy mix: a highly competitive nuclear fleet, the vital resource of fossil-fuelled plants, a proactive approach to renewable energies); a global commitment to corporate social responsibility (human resources and partnerships). (J.S.)

In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report.

Energy Technology Data Exchange (ETDEWEB)

Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V.

1999-04-15

The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report

Energy Technology Data Exchange (ETDEWEB)

Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V

1999-04-15

The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)