WorldWideScience

Sample records for fuel management pwr

  1. PWR fuel management optimization

    International Nuclear Information System (INIS)

    Dumas, Michel.

    1981-10-01

    This report is aimed to the optimization of the refueling pattern of a nuclear reactor. At the beginning of a reactor cycle a batch of fuel assemblies is available: the physical properties of the assemblies are known: the mathematical problem is to determine the refueling pattern which maximizes the reactivity or which provides the flattest possible power distribution. The state of the core is mathematically characterized by a system of partial derivative equations, its smallest eigenvalue and the associated eigenvector. After a study of the convexity properties of the problem, two algorithms are proposed. The first one exhanges assemblies to improve the starting configurations. The enumeration of the exchanges is limited to the 2 by 2, 3 by 3, 4 by 4 permutations. The second one builds a solution in two steps: in the first step the discrete variables are replaced by continuous variables. The non linear optimization problem obtained is solved by ''the Method of Approximation Programming'' and in the second step, the refuelling pattern which provides the best approximation of the optimal power distribution is searched by a Branch an d Bound Method [fr

  2. The verification of PWR-fuel code for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Surian Pinem; Tagor M Sembiring; Tukiran

    2015-01-01

    In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

  3. 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

  4. Make use of EDF orientations in PWR fuel management

    International Nuclear Information System (INIS)

    Gloaguen, A.

    1989-01-01

    The EDF experience acquired permits to allow the PWR fuel performances and to make use of better management. In this domain low progress can be given considerable financial profits. The industrial and commercial structures, the time constant of the fuel cycle, has for consequence that the electric utilities can take advantage only progressively of the expected profits [fr

  5. 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

  6. CORD, PWR Core Design and Fuel Management

    International Nuclear Information System (INIS)

    Trkov, Andrej

    1996-01-01

    1 - Description of program or function: CORD-2 is intended for core design applications of pressurised water reactors. The main objective was to assemble a core design system which could be used for simple calculations (such as frequently required for fuel management) as well as for accurate calculations (for example, core design after refuelling). 2 - Method of solution: The calculations are performed at the cell level with a lattice code in the supercell approximation to generate the single cell cross sections. Fuel assembly cross section homogenization is done in the diffusion approximation. Global core calculations can be done in the full three-dimensional cartesian geometry. Thermohydraulic feedbacks can be accounted for. The Effective Diffusion Homogenization method is used for generating the homogenized cross sections. 3 - Restrictions on the complexity of the problem: The complexity of the problem is selected by the user, depending on the capacity of his computer

  7. 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

  8. 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

  9. 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

  10. 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

  11. A comparative study of fuel management in PWR reactors

    International Nuclear Information System (INIS)

    Barroso, D.E.G.

    1980-01-01

    A comparitive study of fuel recycling in Pressurized Water Reactors was developed, considering not only the conventional uranium cycle, but also the use of thorium as an alternative. The use of thorium was done by varying its conoentration in the homogeneous mixture with uranium in the fuel from 30% up to 90%. The U-233 produced is incorporated within the isotopic composition of irradiated uranium. Various fractions of irradiated recycled fuel to be reprocessed and recycled was considered. Various alternatives of recycling were outlined and a final comparison in the tests done, is furnished in terms of U 3 O 8 and UTS requirements and approximated costs of fuel cycle stages involved. The recycled fuel is considered to be uniformly distributed in the fuel element rods introduced in the nucleus. The influence of the utilization of thorium was also considered for the development of an optimum fuel cycle, regarding the safeguards against nuclear proliferation when utilizing plutonium. A zero-dimensional cellular model was adopted to represent the reactor and the calculus of microscopic cross-sections for the homogenized cell was done by the computer code LEOPARD. A digital computer program was develped for neutronic and fuel depletion calculus and to simulate the refueling of various cycles. (Author) [pt

  12. 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.)

  13. 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

  14. 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

  15. 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.

  16. Practice and prospect of advanced fuel management and fuel technology application in PWR in China

    International Nuclear Information System (INIS)

    Xiao Min; Zhang Hong; Ma Cang; Bai Chengfei; Zhou Zhou; Wang Lei; Xiao Xiaojun

    2015-01-01

    Since Daya Bay nuclear power plant implemented 18-month refueling strategy in 2001, China has completed a series of innovative fuel management and fuel technology projects, including the Ling Ao Advanced Fuel Management (AFM) project (high-burnup quarter core refueling) and the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. First, this paper gives brief introduction to China's advanced fuel management and fuel technology experience. Second, it introduces practices of the advanced fuel management in China in detail, which mainly focuses on the implementation and progress of the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. Finally, the paper introduces the practices of advanced fuel technology in China and gives the outlook of the future advanced fuel management and fuel technology in this field. (author)

  17. The PWR cores management

    International Nuclear Information System (INIS)

    Barral, J.C.; Rippert, D.; Johner, J.

    2000-01-01

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  18. 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.

  19. 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

  20. The database system for the management of technical documentations of PWR fuel design project using CD-ROM

    International Nuclear Information System (INIS)

    Park, Bong Sik; Lee, Won Jae; Ryu, Jae Kwon; Jo, In Hang; Chang, Jong Hwa.

    1996-12-01

    In this report, the database system developed for the management of technical documentation of PWR fuel design project using CD-ROM (compact disk - read only memory) is described. The database system, KIRDOCM (KAERI Initial and Reload Fuel project technical documentation management), is developed and installed on PC using Visual Foxpro 3.0. Descriptions are focused on the user interface of the KIRDOCM. Introduction addresses the background and concept of the development. The main chapter describes the user requirements, the analysis of computing environment, the design of KIRDOCM, the implementation of the KIRDOCM, user's manual of KIRDOCM and the maintenance of the KIRDOCM for future improvement. The implementation of KIRDOCM system provides the efficiency in the management, maintenance and indexing of the technical documents. And, it is expected that KIRDOCM may be a good reference in applying Visual Foxpro for the development of information management system. (author). 2 tabs., 13 figs., 8 refs

  1. Design and fuel management of PWR cores to optimize the once-through fuel cycle

    International Nuclear Information System (INIS)

    Fujita, E.K.; Driscoll, M.J.; Lanning, D.D.

    1978-08-01

    The once-through fuel cycle has been analyzed to see if there are substantial prospects for improved uranium ore utilization in current light water reactors, with a specific focus on pressurized water reactors. The types of changes which have been examined are: (1) re-optimization of fuel pin diameter and lattice pitch, (2) axial power shaping by enrichment gradation in fresh fuel, (3) use of 6-batch cores with semi-annual refueling, (4) use of 6-batch cores with annual refueling, hence greater extended (approximately doubled) burnup, (5) use of radial reflector assemblies, (6) use of internally heterogeneous cores (simple seed/blanket configurations), (7) use of power/temperature coastdown at the end of life to extend burnup, (8) use of metal or diluted oxide fuel, (9) use of thorium, and (10) use of isotopically separated low sigma/sub a/ cladding material. State-of-the-art LWR computational methods, LEOPARD/PDQ-7/FLARE-G, were used to investigate these modifications

  2. 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)

  3. PWR fuel assembly

    International Nuclear Information System (INIS)

    Yamada, Yuji.

    1995-01-01

    A lower end plug is secured to a lower end of a thimble tube. A bolt-like thimble screw is screw-coupled and fastened to a female screw disposed to the end plug by way of a bushing screw-coupled to a lower nozzle. Then, the thimble screw and the lower nozzle are welded to secure the thimble tube and the lower nozzle. The lower portion of the bushing extends near the lower surface of the lower nozzle. The extended portion is provided with a recess to which a bolt head of the thimble screw is tightly inserted and a seating-face portion against which a seating-face of the bolt head abuts. Then, the extended portion of the bushing and the lower nozzle are spot-welded on the side of the lower surface of the nozzle, to prevent rotation of the bushing. This can easily prevent the rotation of the bushing after adjustment, to simplify the assembling of the fuel assembly. (I.N.)

  4. 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).

  5. PWR-to-PWR fuel cycle model using dry process

    International Nuclear Information System (INIS)

    Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

    2002-03-01

    PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

  6. 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)

  7. Quantification of cost of margin associated with in-core nuclear fuel management for a PWR

    International Nuclear Information System (INIS)

    Kropaczek, D.J.; Turinsky, P.J.

    1989-01-01

    The problem of in-core nuclear fuel management optimization is discussed. The problem is to determine the location of core material, such as the fuel and burnable poisons, so as to minimize (maximize) a stated objective within engineering constraints. Typical objectives include maximization of cycle energy production or discharged fuel exposure, and minimization of power peaking factor or reactor vessel fluence. Constraints include discharge burnup limits and one or more of the possible objectives if not selected as the objective. The optimization problem can be characterized as a large combinatorial problem with nonlinear objective function and constraints, which are likely to be active. The authors have elected to employ the integer Monte Carlo programming method to address this optimization problem because of the just-noted problem characteristics. To evaluate the core physics characteristics as a function of fuel loading pattern, second-order accurate perturbation theory is employed with successive application to improve estimates of the optimum loading pattern. No constraints on fuel movement other than requiring quarter-core symmetry were imposed. In this paper the authors employed this methodology to address a related problem. The problem being addressed can be stated as What is the cost associated with margin? Specifically, they wish to assign some financial value in terms of increased levelized fuel cycle cost associated with an increase in core margin of some type, such as power peaking factor

  8. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    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 BWR and PWR. 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 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  9. Multi level optimization of burnable poison utilization for advanced PWR fuel management

    Science.gov (United States)

    Yilmaz, Serkan

    The objective of this study was to develop an unique methodology and a practical tool for designing burnable poison (BP) pattern for a given PWR core. Two techniques were studied in developing this tool. First, the deterministic technique called Modified Power Shape Forced Diffusion (MPSFD) method followed by a fine tuning algorithm, based on some heuristic rules, was developed to achieve this goal. Second, an efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem for a reference Three Mile Island-1 (TMI-1) core. This thesis presents the step by step progress in developing such a tool. The developed deterministic method appeared to perform as expected. The GA technique produced excellent BP designs. It was discovered that the Beginning of Cycle (BOC) Kinf of a BP fuel assembly (FA) design is a good filter to eliminate invalid BP designs created during the optimization process. By eliminating all BP designs having BOC Kinf above a set limit, the computational time was greatly reduced since the evaluation process with reactor physics calculations for an invalid solution is canceled. Moreover, the GA was applied to develop the BP loading pattern to minimize the total Gadolinium (Gd) amount in the core together with the residual binding at End-of-Cycle (EOC) and to keep the maximum peak pin power during core depletion and Soluble boron concentration at BOC both less than their limit values. The number of UO2/Gd2O3 pins and Gd 2O3 concentrations for each fresh fuel location in the core are the decision variables and the total amount of the Gd in the core and maximum peak pin power during core depletion are in the fitness functions. The use of different fitness function definition and forcing the solution movement towards to desired region in the solution space accelerated the GA runs. Special emphasize is given to minimizing the residual binding to increase core lifetime as

  10. INSIGHT: an integrated scoping analysis tool for in-core fuel management of PWR

    International Nuclear Information System (INIS)

    Yamamoto, Akio; Noda, Hidefumi; Ito, Nobuaki; Maruyama, Taiji.

    1997-01-01

    An integrated software tool for scoping analysis of in-core fuel management, INSIGHT, has been developed to automate the scoping analysis and to improve the fuel cycle cost using advanced optimization techniques. INSIGHT is an interactive software tool executed on UNIX based workstations that is equipped with an X-window system. INSIGHT incorporates the GALLOP loading pattern (LP) optimization module that utilizes hybrid genetic algorithms, the PATMAKER interactive LP design module, the MCA multicycle analysis module, an integrated database, and other utilities. Two benchmark problems were analyzed to confirm the key capabilities of INSIGHT: LP optimization and multicycle analysis. The first was the single cycle LP optimization problem that included various constraints. The second one was the multicycle LP optimization problem that includes the assembly burnup limitation at rod cluster control (RCC) positions. The results for these problems showed the feasibility of INSIGHT for the practical scoping analysis, whose work almost consists of LP generation and multicycle analysis. (author)

  11. 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

  12. 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.

  13. 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)

  14. 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.

  15. 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

  16. Preliminary study on direct recycling of spent PWR fuel in PWR system

    International Nuclear Information System (INIS)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  17. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

    1980-01-01

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  18. Advanced PWR fuel design concepts

    International Nuclear Information System (INIS)

    Andersor, C.K.; Harris, R.P.; Crump, M.W.; Fuhrman, N.

    1987-01-01

    For nearly 15 years, Combustion Engineering has provided pressurized water reactor fuel with the features most suppliers are now introducing in their advanced fuel designs. Zircaloy grids, removable upper end fittings, large fission gas plenum, high burnup, integral burnable poisons and sophisticated analytical methods are all features of C-E standard fuel which have been well proven by reactor performance. C-E's next generation fuel for pressurized water reactors features 24-month operating cycles, optimal lattice burnable poisons, increased resistance to common industry fuel rod failure mechanisms, and hardware and methodology for operating margin improvements. Application of these various improvements offer continued improvement in fuel cycle economics, plant operation and maintenance. (author)

  19. 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

  20. 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

  1. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    Provost, Jean-Luc; Debes, Michel

    2005-01-01

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

  2. 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)

  3. 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

  4. PWR and WWER fuel performance. A comparison of major characteristics

    International Nuclear Information System (INIS)

    Weidinger, H.

    2006-01-01

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  5. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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 [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    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 {sup 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).

  7. 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)

  8. 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

  9. 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)

  10. A non-algorithmic approach to the In-core-fuel management problem of a PWR core

    International Nuclear Information System (INIS)

    Kimhy, Y.

    1992-03-01

    The primary objective of a commercial nuclear power plant operation is to produce electricity a low cost while satisfying safety constraints imposed on the operating conditions. Design of a fuel reload cycle for the current generation nuclear power plant represents a multistage process with a series of design decisions taken at various time points. Of these stages, reload core design is an important stage, due to its impact on safety and economic plant performance parameters. Overall. performance of the plant during the power production cycle depends on chosen fresh fuel parameters, as well as specific fuel configuration of the reactor core. The motivation to computerize generation and optimization of fuel reload configurations follows from some reasons: first, reload is performed periodically and requires manipulation of a large amount of data. second, in recent years, more complicated fuel loading patterns were developed and implemented following changes in fuel design and/or operational requirements, such as, longer cycles, advanced burnable poison designs, low leakage loading patterns and reduction of irradiation-induced damage of the pressure vessel. An algorithmic approach to the problem was generally adopted. The nature of the reload design process is a 'heuristic' search performed manually by a fuel manager. The knowledge used by the fuel manager is mostly accumulated experience in reactor physics and core calculations. These features of the problem and the inherent disadvantage of the algorithmic method are the main reasons to explore a non-algorithmic approach for solving the reload configuration problem. Several features of the 'solutions space' ( a collection of acceptable final configurations ) are emphasized in this work: 1) the space contain numerous number of entities (> 25) that are distributed un homogeneously, 2) the lack of a monotonic objective function decrease the probability to find an isolated optimum configuration by depth first search or

  11. Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

    International Nuclear Information System (INIS)

    Sayyaadi, Hoseyn; Sabzaligol, Tooraj

    2010-01-01

    A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

  12. 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

  13. 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)

  14. 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)

  15. 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

  16. Reactor core design calculations and fuel management in PWR; Izracun projekta sredice in upravljanja z forivom tlacnovodnega reaktorja

    Energy Technology Data Exchange (ETDEWEB)

    Ravnik, M [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1987-07-01

    Computer programs and methods developed at J. Stefan Institute for nuclear core design of Krsko NPP are treated. development, scope, verification and organisation of core design procedure are presented. The core design procedure is applicable to any NPP of PWR type. (author)

  17. 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

  18. 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.

  19. 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

  20. 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

  1. PWR in-core nuclear fuel management optimization utilizing nodal (non-linear NEM) generalized perturbation theory

    International Nuclear Information System (INIS)

    Maldonado, G.I.; Turinsky, P.J.; Kropaczek, D.J.

    1993-01-01

    The computational capability of efficiently and accurately evaluate reactor core attributes (i.e., k eff and power distributions as a function of cycle burnup) utilizing a second-order accurate advanced nodal Generalized Perturbation Theory (GPT) model has been developed. The GPT model is derived from the forward non-linear iterative Nodal Expansion Method (NEM) strategy, thereby extending its inherent savings in memory storage and high computational efficiency to also encompass GPT via the preservation of the finite-difference matrix structure. The above development was easily implemented into the existing coarse-mesh finite-difference GPT-based in-core fuel management optimization code FORMOSA-P, thus combining the proven robustness of its adaptive Simulated Annealing (SA) multiple-objective optimization algorithm with a high-fidelity NEM GPT neutronics model to produce a powerful computational tool used to generate families of near-optimum loading patterns for PWRs. (orig.)

  2. 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)

  3. 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)

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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)

  12. 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.).

  13. 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)

  14. Modified ADS molten salt processes for back-end fuel cycle of PWR spent fuel

    International Nuclear Information System (INIS)

    Choi, In-Kyu; Yeon, Jei-Won; Kim, Won-Ho

    2002-01-01

    The back-end fuel cycle concept for PWR spent fuel is explained. This concept is adequate for Korea, which has operated both PWR and CANDU reactors. Molten salt processes for accelerator driven system (ADS) were modified both for the transmutation of long-lived radioisotopes and for the utilisation of the remained fissile uranium in PWR spent fuels. Prior to applying molten salt processes to PWR fuel, hydrofluorination and fluorination processes are applied to obtain uranium hexafluoride from the spent fuel pellet. It is converted to uranium dioxide and fabricated into CANDU fuel. From the remained fluoride compounds, transuranium elements can be separated by the molten salt technology such as electrowinning and reductive extraction processes for transmutation purpose without weakening the proliferation resistance of molten salt technology. The proposed fuel cycle concept using fluorination processes is thought to be adequate for our nuclear program and can replace DUPIC (Direct Use of spent PWR fuel in CANDU reactor) fuel cycle. Each process for the proposed fuel cycle concept was evaluated in detail

  15. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched 235U fuel pins

    International Nuclear Information System (INIS)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched 235 U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are rather

  16. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

  17. 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

  18. Assessment of cold composite fuels for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Coulon-Picard, E.; Agard, M.; Boulore, A.; Castelier, E.; Chabert, C.; Conti, A.; Frayssines, P.E.; Lechelle, J.; Maillard, S.; Matheron, P.; Pelletier, M.; Phelip, M.; Piluso, P.; Vaudano, A

    2009-06-15

    This study is devoted to evaluation of a new innovative micro structured fuel for future pressurized water reactor. This fuel would have potential to increase the safety margins, lowering fuel temperatures by adding a small fraction of a high conductivity second phase material in the oxide fuel phase. The behavior of this fuel in a standard rod has been modeled with finite element codes and was assessed for different aspects of the cycle as neutronic studies, thermal behavior, reprocessing and economics. Feasibility of fuels has been investigated with the fabrication and characterizations of the microstructure of composite fuels with powder metallurgy and HIP processes. First, a CERCER (Ceramic = UO{sub 2}- Ceramic matrix made of silicon carbide, SiC) fuel type has been investigated, the advantages of a ceramic being generally its transparency to neutrons and its high melting temperature. A first design of kernel type fuel was first chosen with a gap between the UO{sub 2} particles and the second phase material in order to avoid mechanical interaction between the two components. Due to lowering thermal conductivity of SiC under irradiation, this CERCER fuel did not allow a temperature gain compared to current fuel. No ceramic material seems to exhibit all required properties. Even beryllium oxide (BeO), which conductivity does not decrease with irradiation according to the literature, induces difficulties with ({alpha}, n) reactions and toxicity. The study then focused on Cermet fuels (Ceramic-Metal). The metal matrix must be transparent to neutrons and have a good thermal conductivity. Several materials have been considered such as zirconium alloys, austenitic and ferritic stainless steals and chromium based alloys. The heterogeneous composite fuels were modeled using the 3D/CASTM finite element code. From an economical and neutron point of view, it was important to keep a low fraction of metal phase, i.e. less than 10 % of Zr for example. However, the fuel

  19. 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

  20. 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)

  1. Timing analysis of PWR fuel pin failures

    International Nuclear Information System (INIS)

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J.; Straka, M.

    1992-09-01

    Research has been conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin bumup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PFL/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B ampersand W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B ampersand W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design bumup. Using peaking factors commensurate widi actual bumups would result in longer intervals for both reactor designs. This document also contains appendices A through J of this report

  2. 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)

  3. 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

  4. 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)

  5. 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.)

  6. PWR fuel rod corrosion in Japan

    International Nuclear Information System (INIS)

    Inoue, S.; Mori, K.; Murata, K.; Kobasyashi, S.

    1997-01-01

    Many particular appearance were observed on the fuel rod surfaces during fuel inspection at reactor outage in 1991. The appearances looked like small black circular nodules. The size was approximately 1 mm. This kind of appearances were found on fuel rods of which burnup exceeded approximately 30 GWd/t and at the second or third spans of the fuel assembly from the top. In order to clarify the cause, PIE was performed. The black nodules were confirmed to be oxide film spalling by visual inspection. Maximum oxide film thickness was 70 μm and spalling was observed where oxide thickness exceeded 40 t0 50 μm. Oxide film thickness was greater than expected. Many small pores were found in the oxide film when the oxide film had become thicker. Many circumferential cracks were also found in the film. It was speculated that these cracks caused the spalling of the oxide film. Hydride precipitates were mainly oriented circumferentially. Dense hydrides were observed near the outer rim of the cladding. No concentrated hydrides were observed near the spalling area. Maximum hydrogen content was 315 ppm. It was confirmed that the results of tensile test showed no significant effects by corrosion. The mechanism of accelerated corrosion was studied in detail. Water chemistry during irradiation was examined. Lithium content was maintained below 2.2 ppm. pH value was kept between 6.9 and 7.2. There was no anomalies in water chemistry during reactor operation. Cladding fabrication record clarified that heat treatment parameter was smaller than the optimum value. In Japan, heat treatment of the cladding was already optimized by improved fabrication process. Also chemical composition optimization of the cladding, such as low Tin and high Silicon content, was adopted for high burnup fuel. These remedies has already reduced fuel cladding corrosion and we believe we have solved this problem. (author). 6 figs, 1 tab

  7. PWR fuel rod corrosion in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Inoue, S [Kansai Electric Power Co., Inc., Osaka (Japan); Mori, K; Murata, K; Kobasyashi, S [Nuclear Fuel Industries, Ltd, Osaka (Japan)

    1997-02-01

    Many particular appearance were observed on the fuel rod surfaces during fuel inspection at reactor outage in 1991. The appearances looked like small black circular nodules. The size was approximately 1 mm. This kind of appearances were found on fuel rods of which burnup exceeded approximately 30 GWd/t and at the second or third spans of the fuel assembly from the top. In order to clarify the cause, PIE was performed. The black nodules were confirmed to be oxide film spalling by visual inspection. Maximum oxide film thickness was 70 {mu}m and spalling was observed where oxide thickness exceeded 40 t0 50 {mu}m. Oxide film thickness was greater than expected. Many small pores were found in the oxide film when the oxide film had become thicker. Many circumferential cracks were also found in the film. It was speculated that these cracks caused the spalling of the oxide film. Hydride precipitates were mainly oriented circumferentially. Dense hydrides were observed near the outer rim of the cladding. No concentrated hydrides were observed near the spalling area. Maximum hydrogen content was 315 ppm. It was confirmed that the results of tensile test showed no significant effects by corrosion. The mechanism of accelerated corrosion was studied in detail. Water chemistry during irradiation was examined. Lithium content was maintained below 2.2 ppm. pH value was kept between 6.9 and 7.2. There was no anomalies in water chemistry during reactor operation. Cladding fabrication record clarified that heat treatment parameter was smaller than the optimum value. In Japan, heat treatment of the cladding was already optimized by improved fabrication process. Also chemical composition optimization of the cladding, such as low Tin and high Silicon content, was adopted for high burnup fuel. These remedies has already reduced fuel cladding corrosion and we believe we have solved this problem. (author). 6 figs, 1 tab.

  8. 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)

  9. 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)

  10. 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

  11. 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

  12. 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

  13. Strategy of fuel management

    International Nuclear Information System (INIS)

    Guesdon, B.; Le Bars, M.; Mathonniere, G.

    1996-01-01

    The management of nuclear fuels in PWR type reactors has been adapted to improve the safety and the competitiveness of brackets. The economic optimum, at the park level, depends on many parameters, variable with time and in function of them, we favour the annual campaigns and the economy won on the cost of cycle, or long campaigns with benefit on availability. The reduction of the number of stopping improves the availability, limits the doses integrated by the personnel of intervention and reduces the number of incidents during the stopping. An other determining factor is connected to the policy of closed cycle with the the principle of equality between the reprocessing flux and the valorization of reprocessed fuels: plutonium and reprocessed uranium. The progress of fuel have allowed significant improvements in the managements of cores. With the safety, the aim is also to keep if not improve the competitiveness of the Nuclear park by valorizing the matter coming from reprocessing. (N.C.)

  14. 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.)

  15. 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

  16. 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.)

  17. 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)

  18. 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)

  19. 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

  20. 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.

  1. Plutonium recycle. In-core fuel management

    International Nuclear Information System (INIS)

    Vincent, F.; Berthet, A.; Le Bars, M.

    1985-01-01

    Plutonium recycle in France will concern a dozen of PWR 900 MWe controlled in gray mode till 1995. This paper presents the main characteristics of fuel management with plutonium recycle. The organization of management studies will be copied from this developed for classical management studies. Up these studies, a ''feasibility report'' aims at establishing at each stage of the fuel cycle, the impact of the utilization of fuel containing plutonium [fr

  2. 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

  3. 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)

  4. 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

  5. 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

  6. 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)

  7. Spent fuel management in France: Programme status

    International Nuclear Information System (INIS)

    Chaudat, J.P.

    1990-01-01

    France's programme is best characterized as a closed fuel cycle including reprocessing, Plutonium recycling in PWR and use of breeder reactors. The current installed nuclear capacity is 52.5 GWe from 55 units. The spent fuel management scheme chosen is reprocessing. This paper describes the national programme, spent nuclear fuel storage, reprocessing and contracts for reprocessing of spent fuel from various countries. (author). 5 figs, 2 tabs

  8. 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)

  9. DUPIC fuel fabrication using spent PWR fuels at KAERI

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-12-01

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

  10. In-core fuel management activities in China

    International Nuclear Information System (INIS)

    Ruan Keqiang; Chen Renji; Hu Chuanwen

    1990-01-01

    The development of nuclear power in China has reached such a stage that PWR in-core fuel management becomes an urgent problem. At present the main effort is concentrated on solving the Qinshan nuclear power plant and Daya Bay nuclear power plant fuel management problems. For the Qinshan PWR (300 MWe) two packages of in-core fuel management code were developed, one with simplified nodal diffusion method and the other uses advanced Green's function nodal method. Both were used in the PWR core design. With the help of the two code packages first two cycles of the Qinshan PWR core burn-up were calculated. Besides, several research works are under way in the following areas: improvement of the nodal diffusion method and other coarse mesh method in terms of computing speed and accuracy; backward diffusion technique for fuel management application; optimization technique in the fuel loading pattern searching. As for the Daya Bay PWR plant (twin 900 MWe unit), the problem about using what kind of code package for in-core fuel management is still under discussion. In principle the above mentioned code packages are also applicable to it. Besides PWR, in-core fuel management research works are also under way for research reactors, for example, heavy water research reactor and high flux research reactor in some institutes in China. China also takes active participation in international in-core fuel management activities. (author). 19 refs

  11. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  12. 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

  13. Core management and performance analysis for PWR

    International Nuclear Information System (INIS)

    Lee, J.B.; Lee, C.K.; Kim, J.S.; Lee, S.K.; Moon, K.S.; Chun, B.J.; Chang, J.W.; Kim, Y.J.

    1981-01-01

    The KINS (KAERI Improved Nodal Simulation) program, a three-dimensional nodal simulation code for pressurized water reactor fuel management, has been developed and benchmarked against the cycles 1 and 2 of the Kori-1 reactor. The critical boron concentration and three-dimensional power distribution at BOL, HZP condition have been calculated and compared with the operating data. A three-dimensional depletion calculation at HFP condition has been performed for cycle 1 with an interval of 1000 MWD/MTU and compared with the operating data. Similar calculation was also performed for cycle 2 and then compared with the design data of the reactor vendor. At the same time, a prediction of in-core detectors reaction rate was made so as to be compared with the operating data. As the result of comparisons, our calculation as well as the justification of the correlations is shown to be in excellent agreement with the operating data within an allowable limit

  14. Seismic behaviour of PWR fuel assemblies model and its validation

    International Nuclear Information System (INIS)

    Queval, J.C.; Gantenbein, F.; Brochard, D.; Benjedidia, A.

    1991-01-01

    The validity of the models simulating the seismic behaviour of PWR cores can only be exactly demonstrated by seismic testing on groups of fuel assemblies. Shake table seismic tests of rows of assembly mock-ups, conducted by the CEA in conjunction with FRAMATOME, are presented in reference /1/. This paper addresses the initial comparisons between model and test results for a row of five assemblies in air. Two models are used: a model with a single beam per assembly, used regularly in accident analyses, and described in reference /2/, and a more refined 2-beam per assembly model, geared mainly towards interpretation of test results. The 2-beam model is discussed first, together with parametric studies used to characterize it, and the study of the assembly row for a period limited to 2 seconds and for different excitation levels. For the 1-beam model assembly used in applications, the row is studied over the total test time, i.e twenty seconds, which covers the average duration of the core seismic behaviour studies, and for a peak exciting acceleration value at 0.4 g, which corresponds to the SSE level of the reference spectrum

  15. 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)

  16. 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)

  17. 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

  18. 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

  19. 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

  20. Life management plants at nuclear power plants PWR

    International Nuclear Information System (INIS)

    Esteban, G.

    2014-01-01

    Since in 2009 the CSN published the Safety Instruction IS-22 (1) which established the regulatory framework the Spanish nuclear power plants must meet in regard to Life Management, most of Spanish nuclear plants began a process of convergence of their Life Management Plants to practice 10 CFR 54 (2), which is the current standard of Spanish nuclear industry for Ageing Management, either during the design lifetime of the plant, as well as for Long-Term Operation. This article describe how Life Management Plans are being implemented in Spanish PWR NPP. (Author)

  1. Accelerated fuel depreciation as an economic incentive for low-leakage fuel management

    International Nuclear Information System (INIS)

    Downar, T.J.

    1986-01-01

    An analysis is presented which evaluates the tax depreciation advantage which results from the increased rate of fuel depletion achieved in the current low-leakage fuel-management LWR core reload designs. An analytical fuel-cycle cost model is used to examine the important cost parameters which are then validated using the fuel-cycle cost code CINCAS and data from the Maine Yankee PWR. Results show that low-leakage fuel management, through the tax depreciation advantage from accelerated fuel depletion, provides an improvement of several percent in fuel-cycle costs compared to traditional out-in fuel management and a constant fuel depletion rate. (author)

  2. 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.

  3. 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

  4. 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

  5. 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)

  6. Fuel management

    International Nuclear Information System (INIS)

    Schwarz, E.R.

    1975-01-01

    Description of the operation of power plants and the respective procurement of fuel to fulfil the needs of the grid. The operation of the plants shall be optimised with respect to the fuel cost. (orig./RW) [de

  7. PWR accident management realated tests: some Bethsy results

    International Nuclear Information System (INIS)

    Clement, P.; Chataing, T.; Deruaz, R.

    1993-01-01

    The BETHSY integral test facility which is a scaled down model of a 3 loop FRAMATOME PWR and is currently operated at the Nuclear Center of Grenoble, forms an important part of the French strategy for PWR Accident Management. In this paper the features of both the facility and the experimental program are presented. Two accident transients: a total loss of feedwater and a 2'' cold leg break in case of High Pressure Safety Injection System failure, involving either Event Oriented - or State Oriented-Emergency Operating Procedures (EO-EOP or SO-EOP) are described and the system response analyzed. CATHARE calculation results are also presented which illustrate the ability of this code to adequately predict the key phenomena of these transients. (authors). 13 figs., 11 refs., 2 tabs

  8. 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)

  9. 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

  10. 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

  11. 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

  12. Benchmark solution of contemporary PWR integral fuel burnable absorbers

    International Nuclear Information System (INIS)

    Stucker, D.L.; Hone, M.J.; Holland, R.A.

    1993-01-01

    This paper presents a closely controlled benchmark solution of the two major contemporary pressurized water reactor integral burnable absorber designs: zirconium diboride (ZrB 2 ) and gadolinia (Gd 2 O 3 ). The comparison is accomplished using self-generating equilibrium cycles with equal energy, equal discharge burnup, and equal safety constraints. The reference plant for this evaluation is a 3411-MW(thermal) Westinghouse four-loop nuclear steam supply system operating with an inlet temperature of 285.9 degrees C, a core coolant mass now rate of 16877.3 kg/s, and coolant pressure of 15.5 MPa. The reactor consists of 193 VANTAGE 5H fuel assemblies that are discharged at a region average burnup of 48.4 GWd/tonne U. Each fuel assembly contains a natural uranium axial blanket 15.24 cm long at the top and the bottom of the fuel rod. The burnable absorber rods are symmetrically radially dispersed within the fuel assembly such that intrabundle power peaking is minimized. The burnable absorber material for both ZrB 2 and Gd 2 O 3 is axially zoned to the central 304.8 cm of the absorber-bearing fuel rods. The fuel management was constrained such that the thermal and safety limitations of F δH q -5 /degrees C were simultaneously achieved. The maximum long-term operating soluble boron concentration was also limited to 446 effective full-power days (EFPDs) including 14 EFPDs of power coastdown were assumed

  13. Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

    International Nuclear Information System (INIS)

    Wati, Nurokhim

    2008-01-01

    Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

  14. 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.

  15. 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)

  16. 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

  17. 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

  18. 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

  19. 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.

  20. Thorium fuel cycle management

    International Nuclear Information System (INIS)

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

    2010-01-01

    In this presentation author deals with the thorium fuel cycle management. Description of the thorium fuels and thorium fuel cycle benefits and challenges as well as thorium fuel calculations performed by the computer code HELIOS are presented.

  1. 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

  2. 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)

  3. US PWR steam generator management: An overview

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.

    1997-01-01

    This paper provides an overview on the status of steam generator management activities in US PWRs, and includes: (1) an overview of the impact of steam generator problems; (2) a brief discussion of historical damage trends and the current damage mechanism of most concern; (3) a discussion of the elements of open-quotes steam generator managementclose quotes; and (4) a description of the approach being followed to implement a degradation-specific protocol for tubing inspection and repair. This paper was prepared in conjunction with another paper presented during the Plenary Session of this Conference, open-quotes Steam Generator Degradation: Current Mitigation Strategies for Controlling Corrosionclose quotes, and is provided as a supplement to that material

  4. 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

  5. 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

  6. 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)

  7. 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.

  8. 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.

  9. 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

  10. Radioactive waste management of experimental DUPIC fuel fabrication process

    International Nuclear Information System (INIS)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Hong, K. P.

    2001-01-01

    The concept of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) is a dry processing technology to manufacture CANDU compatible DUPIC fuel from spent PWR fuel material. Real spent PWR fuel was used in IMEF M6 hot cell to carry out DUPIC experiment. Afterwards, about 200 kg-U of spent PWR fuel is supposed to be used till 2006. This study has been conducted in some hot cells of PIEF and M6 cell of IMEF. There are various forms of nuclear material such as rod cut, powder, green pellet, sintered pellet, fabrication debris, fuel rod, fuel bundle, sample, and process waste produced from various manufacturing experiment of DUPIC fuel. After completing test, the above nuclear wastes and test equipment etc. will be classified as radioactive waste, transferred to storage facility and managed rigorously according to domestic and international laws until the final management policy is determined. It is desirable to review management options in advance for radioactive waste generated from manufacturing experiment of DUPIC nuclear fuel as well as residual nuclear material and dismantled equipment. This paper includes basic plan for DUPIC radwaste, arising source and estimated amount of radioactive waste, waste classification and packing, transport cask, transport procedures

  11. 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)

  12. 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.

  13. 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)

  14. 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)

  15. 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 PWR for use in the naval reactor programme and the subsequent construction and operation of the nuclear power plant at Shippingport in 1957. The development of the BWR in 1954 and its selection for the plant at Dresden in 1959 established this concept as the other major reactor type in the US nuclear power programme. The subsequent growth profile is presented. A significant operating record has been accumulated concerning the availability of each of these reactor types. In addition, the use and performance of BWRs and PWRs 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 ensure effective safeguards at nuclear power installations; current measures are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. Both fuel cycles are examined in terms of: fuel burnup experience and prospects for improvement; natural uranium resources; enrichment capacity; 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 in terms of available capacity and plans for expansion. The US electric utility industry viewpoint regarding an ultimate programme for waste management is outlined. Finally, the current economics and future cost trends of nuclear power plants are evaluated. (author)

  16. Siemens advance PWR fuel assemblies (HTP) and cladding

    International Nuclear Information System (INIS)

    Stout, R. B.; Woods, K. N.

    1997-01-01

    This paper describes the key features of the Siemens HTP (High Thermal Performance) fuel design, the current in-reactor performance of this advanced fuel assembly design, and the advanced cladding types available

  17. Development of failed fuel detection system for PWR (III)

    International Nuclear Information System (INIS)

    Hwang, Churl Kew; Kang, Hee Dong; Jeong, Seung Ho; Cho, Byung Sub; Yoon, Byeong Joo; Yoon, Jae Seong

    1987-12-01

    Ultrasonic transducers satisfying the conditions for failed fuel rod detection for failed fuel rod detection have been designed and built. And performance tests for them have been carried out. Ultrasonic signal processing units, a manipulator guiding the ultrasonic probe through the fuel assembly lanes and its control units have been constructed. The performance of the system has been verified experimentally to be successful in failed fuel rod detection. (Author)

  18. Technical report: fabrication of PWR type rodlet fuel

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Uno, Hisao; Sasajima, Hideo

    1990-06-01

    With respect to the simulated reactivity initiated accident (RIA) experiments with pre-irradiated LWR type fuel rods at nuclear safety research reactor (NSRR), there were principally three technical difficulties which should be overcome: (1) Fabrication of the rodlet fuel; Fuel rods from the commercial power reactors had an active column length by 3.6m. To utilize this for NSRR pulse experiment, rodlet fuel having an active column length by 0.12m (reduced to one thirtieth) is requested to fabricate without changing the inside fuel conditions. (2) Development of in-core instrumentations: During pre-irradiation stages, a long-sized fuel rod had dimensional changes by waterside corrosion, bowing, creep down and so on. The fuel also had greater amount of radioactive fission products. This condition is significant to in-core instrumentations to be attached to the fuel rods. Well characterized data to be obtained from these, however, are quite necessary and important from research point of view. Remote handling techniques to attach the rod pressure sensor, the cladding extensometer, the fuel extensometer, and the cladding surface thermocouple to pre-irradiated fuel rods are, therefore, requested to develop. (3) Installation of PIE equipments for pulsed rodlet fuels: PIE on the pulsed rodlet fuels are necessary to better understanding the fuel performance detaily. Equipments which can easily detect the data related to PCMI type fuel failure are matter of concern. Since 1986, the technical difficulties have been tried to overcome by all staffs belonging to Reactivity Accident Laboratory, NSRR Operation Division, Department of Reactor Fuel Examination and Hot Laboratory. This report describes the technical achievements obtained through four years work. (author)

  19. 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)

  20. 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)

  1. Experience of Areva in fuel services for PWR and BWR

    International Nuclear Information System (INIS)

    Morales, I.

    2015-01-01

    AREVA being an integrated supplier of fuel assemblies has included in its strategy to develop services and solutions to customers who desire to improve the performance and safety of their fuel. These services go beyond the simple 'after sale' services that can be expected from a fuel supplier: The portfolio of AREVA includes a wide variety of services, from scientific calculations to fuel handling services in a nuclear power plant. AREVA is committed to collaborate and to propose best-in-class solutions that really make the difference for the customer, based on 40 years of Fuel design and manufacturing experience. (Author)

  2. Further developments of PWR and BWR fuel elements

    International Nuclear Information System (INIS)

    Sofer, G.A.; Busselman, G.J.; Federico, L.J.

    1988-01-01

    The performance, safety, and economy of nuclear power plants in inluenced very decisively by the quality of their fuel elements. This is why quality assurance in fuel fabrication has been a factor of great importance from the outset. Operating experince and more stringent performance requirements have resulted in a continuous process of further development of fuel elements, which has been reflected also in lower and lower failure rates and increasingly higher burn-ups. Next to further development also innovation has been an important factor contributing to the present high quality level of fuel elements, which also has allowed fuel cycle costs to be decreased quite considerably. (orig.) [de

  3. 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.

  4. 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

  5. Contribution to the experimental qualification of PWR fuel storage calculations

    International Nuclear Information System (INIS)

    Marsault, Philippe.

    1980-12-01

    Experiments were carried out on assemblies representative of those used in PWR reactors in a configuration made critical with a driver zone. In this way, certain parameters were able to be measured using current classical techniques. As the multiplication factor for a group of assemblies cannot be determined directly, substitutions were made with an equivalent homogeneous lattice in which Laplacian measurements could be made. The k(infinite) factor was obtained by introducing a migration area which can only be obtained from calculations. Experimental storage studies realized during the CRISTO 1 campaign utilize: 1) a lattice with 4 14x14 pin assemblies immersed in ordinary water; 2) a lattice with 4 14x14 pin assemblies and 3) a regular lattice. The CRISTO experiment enabled criticality calculations to be qualified with these lattices for storage under accidental conditions [fr

  6. 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)

  7. 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.

  8. 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

  9. 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)

  10. 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)

  11. 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

  12. 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

  13. 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.)

  14. 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)

  15. 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)

  16. 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

  17. 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

  18. Conceptual development of a test facility for spent fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs.

  19. Conceptual development of a test facility for spent fuel management

    International Nuclear Information System (INIS)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G.

    1997-01-01

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

  20. Waste management aspects of entire PWR LOOP decontamination

    International Nuclear Information System (INIS)

    Murray, A.P.; Roesmer, J.

    1988-01-01

    The waste management parameters for decontamination of an entire PWR primary circuit have been determined for dilute alkaline-permanganate/citric acid (APCA), LOMI, ozone and cerium acid process variations. APCA processes generate the largest waste volumes; over 140 m 3 (5000 ft 3 ) in some cases. This represents a potential disposal cost of one million dollars. The cation regeneration column makes the greatest contribution to the disposal volume. In contrast, the LOMI process generates approximately half as much waste, but it is expected to contain relatively high metal concentrations (200-800 ppm). The ozone and cerium acid processes product the least waste, usually under 45 m 3 . These waste volume estimates represent considerable fractions of a utility's annual disposal volume. Consequently, improved waste processing technology is required, and several approaches are suggested

  1. Characteristic test technology for PWR fuel and its components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho; Jeong, Yong Hwan; Park, Sang Yoon; Kim, Kyeng Ho; Nam, Cheol; Baek, Jong Hyuk; Lee, Myung Ho; Choi, Byoung Kwon; Song, Kun Woo; Kang, Ki Won; Kim, Keon Sik; Kim, Jong Hun; Kim, Young Min; Yang, Jae Ho; Song, Kee Nam; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Chun, Tae Hyun; In, Wang Kee; Oh, Dong Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-01-01

    Characteristic tests of fuel assembly and its components being developed in the Advanced LWR Fuel Development Project supported by the mid-long term nuclear R and D program are described in this report. Performance verification of fuel and its components by the characteristic tests are essential to their development. Fuel components being developed in the Advanced LWR Fuel Development Project are zirconium alloy cladding, UO{sub 2} and burnable absorber pellets, spacer grid and top and bottom end pieces. Detailed test plans for those fuel components are described in this report, and test procedures of cladding and pellet are also described in the Appendix. Examples of the described tests are in- and out-of- pile corrosion and mechanical tests such as creep and burst tests for the cladding, in-pile capsule and ramp tests for the pellet, mechanical tests such as strength and vibration, and thermal-hydraulic tests such as pressure drop and critical heat flux for the spacer grid and top and bottom end pieces. It is expected that this report could be used as the standard reference for the performance verification tests in the development of LWR fuel and its components. 11 refs., 9 figs., 2 tabs. (Author)

  2. 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)

  3. 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.

  4. 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)

  5. The PWR fuel cycle. Utilization of uranium in a reactor

    International Nuclear Information System (INIS)

    Mignot, E.

    After having briefly described the core of a pressurized water reactor, the fuel is examined and, in particular, the change in reactivity that governs the renewal of the fuel. The present French nuclear units are taken as example and it is shown that with the development of the nuclear complex, it is no longer possible to reason on the basis of an isolated reactor, since the running of a reactor is set by the network and its working constraints become a priority. The optimization of the fuel control must therefore cover the total cost [fr

  6. Life management plants at nuclear power plants PWR; Planes de gestion de vida en centrales nucleares PWR

    Energy Technology Data Exchange (ETDEWEB)

    Esteban, G.

    2014-10-01

    Since in 2009 the CSN published the Safety Instruction IS-22 (1) which established the regulatory framework the Spanish nuclear power plants must meet in regard to Life Management, most of Spanish nuclear plants began a process of convergence of their Life Management Plants to practice 10 CFR 54 (2), which is the current standard of Spanish nuclear industry for Ageing Management, either during the design lifetime of the plant, as well as for Long-Term Operation. This article describe how Life Management Plans are being implemented in Spanish PWR NPP. (Author)

  7. 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.)

  8. 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.

  9. Aspects regarding the fuel management for PHWR nuclear reactors

    International Nuclear Information System (INIS)

    Dragusin, O.; Bobolea, A.; Voicu, A.

    2001-01-01

    Fuel management for PHWR nuclear reactors is completely different from the PWR reactors fuel management. PHWR reactor fuel loading procedures are repeated after an interval of time, as defined and specified in the project documentation, using a fuel machine that can be attached to the terminal fittings of horizontal pressure tubes while the reactor is a full power. Another aspect of fuel management policy is related to the possibility of bi-directional loading of the reactor, with the primary advantage of uniform and symmetrical characteristics. (authors)

  10. The Gd-isotopic fuel for high burnup in PWR's

    International Nuclear Information System (INIS)

    Dias, Marcio Soares; Mattos, João Roberto L. de; Andrade, Edison Pereira de

    2017-01-01

    Today, the discussion about the high burnup fuel is beyond the current fuel enrichment licensing and burnup limits. Licensing issues and material/design developments are again key features in further development of the LWR fuel design. Nevertheless, technological and economical solutions are already available or will be available in a short time. In order to prevent the growth of the technological gap, Brazil's nuclear sector needs to invest in the training of new human resources, in the access to international databases, and in the upgrading existing infrastructure. Experimental database and R&D infrastructure are essential components to support the autonomous development of Brazilian Nuclear Reactors, promoting the development of national technologies. The (U,Gd)O_2 isotopic fuel proposed by the CDTN's staff solve two main issues in the high burnup fuel, which are (1) the peak of reactivity resulting from the Gd-157 fast burnup, and (2) the peak of temperature in the (U,Gd)O_2 nuclear fuel resulting from detrimental effects in the thermal properties for gadolinia additions higher than 2%. A sustainable future can be envisaged for the nuclear energy. (author)

  11. 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)

  12. NDE of PWR fuel: Identifying candidates for hot cell examination

    International Nuclear Information System (INIS)

    Moon, J.E.; Bury, J.G.; Correal, O.A.; Kunishi, H.; Wilson, H.W.

    1992-05-01

    On-site examinations were performed at the Indian Point 3 and Callaway reactors to attempt to identify the leakage mechanism of several leaking fuel rods. The exams consisted of removing the leaking fuel rods from the assembly and performing a visual examination. These results, combined with other available on-site data on leaking fuel rods, were used to select fuel rods for shipment to a hot cell for detailed root cause examination. Three fuel rods from the Indian Point 3 reactor were found to be leaking due to debris-induced fretting. The examinations at Callaway were terminated prior to completion due to utility scheduler conflicts. Rods from the Callaway reactor were selected for shipment to the hot cell along with the rods from the Byron 1 and 2 and V.C. Summer reactors. The data presented in the report summarize the coolant activity history, the UT examination results, and a summary of the review of the fabrication records. The basis for the selection of the rods to be sent to the hot cells is also summarized

  13. The fractalline properties of experimentally simulated PWR fuel crud

    Science.gov (United States)

    Dumnernchanvanit, I.; Mishra, V. K.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Mota, G.; Hussey, D.; Wang, G.; Byers, W. A.; Short, M. P.

    2018-02-01

    The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud may cause serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each makes critical assumptions regarding how to treat the complex, porous microstructure of crud and its resultant effects on temperature, pressure, and crud chemistry. In this study, we demonstrate that crud is indeed a fractalline porous medium using flowing loop experiments, validating the most recent models of its effects on LWR fuel cladding. This crud is shown to match that in other LWR-prototypical facilities through a porosity-fractal dimension scaling law. Implications of this result range from post-mortem analysis of the effects of crud on reactor fuel performance, to utilizing crud's fractalline dimensions to quantify the effectiveness of anti-fouling measures.

  14. FAMREC, PWR Lateral Mechanical Fuel Rod Assembly Response

    International Nuclear Information System (INIS)

    Guenzler, R.C.

    1995-01-01

    1 - Description of program or function: The Fuel Assembly Mechanical Response Code (FAMREC) calculates the lateral mechanical response of a row of fuel assemblies while allowing for two types of nonlinearities. The first type is a geometric nonlinearity in the form of gaps between individual assemblies and between peripheral assemblies and a boundary wall. Impacting is monitored across the gaps. The second nonlinearity is the permanent deformation of the fuel assembly spacer grid to compressive loading. 2 - Method of solution: The response is calculated in the modal plane. The coupled differential equations are solved in closed form using Laplace transformations. The discrete displacements and velocities are then calculated and the gaps in the system monitored at each axial elevation for impacting. These impact forces are then applied statistically at a given time-step, and equilibrium is found using a Gaussian elimination technique. Three impact force calculation methods are available: 1- a linear impact force and crushing load audit calculation, 2- a more detailed linear impact force and crushing load calculation, and 3- a non-linear grid calculation which allows for plastic deformation of the fuel assembly spacer grids. 3 - Restrictions on the complexity of the problem: Maxima of: 3601 time-steps and forces; 80 modes; 30 applied forces; 15 fuel assemblies; and 5 impact grids per assembly

  15. 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

  16. 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

  17. 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)

  18. 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)

  19. 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)

  20. 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).

  1. Zircaloy sheathed thermocouples for PWR fuel rod temperature measurements

    International Nuclear Information System (INIS)

    Anderson, J.V.; Wesley, R.D.; Wilkins, S.C.

    1979-01-01

    Small diameter zircaloy sheathed thermocouples have been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Surface mounted thermocouples were developed to measure the temperature of zircaloy clad fuel rods used in the Thermal Fuels Behavior Program (TFBP), and embedded thermocouples were developed for use by the Loss-of-Fluid Test (LOFT) Program for support tests using zircaloy clad electrically heated nuclear fuel rod simulators. The first objective of this developmental effort was to produce zircaloy sheathed thermocouples to replace titanium sheathed thermocouples and thereby eliminate the long-term corrosion of the titanium-to-zircaloy attachment weld. The second objective was to reduce the sheath diameter to obtain faster thermal response and minimize cladding temperature disturbance due to thermocouple attachment

  2. Modal testing and identification of a PWR fuel assembly

    International Nuclear Information System (INIS)

    Pisapia, S.; Collard, B.; Mori, V.; Bellizzi, S.

    2003-01-01

    This study aims at characterizing the vibratory behavior of a full-scale fuel assembly using an experimental approach. The effect of the assembly environment (air, stagnant water, and water under flow) is studied. The analysis of the test series shows that the vibratory behavior of full-scale fuel assembly is strongly nonlinear. An identification phase, based on temporal mean square criterion, allows us to obtain a nonlinear model representative of the first vibration mode of a fuel assembly. The selected class of models including damping and stiffness nonlinear terms is efficient in air, in stagnant water, and in water under flow. In all environments, the stiffness decreases with the displacement level and the damping increases with the velocity level. In the presence of water, the damping goes up and increases again with flowrate. (author)

  3. Fuel thermal conductivity (FTHCON). Status report. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hagrman, D. L.

    1979-02-01

    An improvement of the fuel thermal conductivity subcode is described which is part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. The original version was published in the Materials Properties (MATPRO) Handbook, Section A-2 (Fuel Thermal Conductivity). The improved version incorporates data which were not included in the previous work and omits some previously used data which are believed to come from cracked specimens. The models for the effect of porosity on thermal conductivity and for the electronic contribution to thermal coductivity have been completely revised in order to place these models on a more mechanistic basis. As a result of modeling improvements the standard error of the model with respect to its data base has been significantly reduced.

  4. Fuel utilization improvements in a once-through PWR fuel cycle. Final report on Task 6

    International Nuclear Information System (INIS)

    Dabby, D.

    1979-06-01

    In studying the position of the United States Department of Energy, Non-proliferation Alternative Systems Assessment Program, this report determines the uranium saving associated with various improvement concepts applicable to a once-through fuel cycle of a standard four-loop Westinghouse Pressurized Water Reactor. Increased discharged fuel burnup from 33,000 to 45,000 MWD/MTM could achieve a 12% U 3 O 8 saving by 1990. Improved fuel management schemes combined with coastdown to 60% power, could result in U 3 O 8 savings of 6%

  5. 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

  6. Modular approach to LWR in-core fuel management

    International Nuclear Information System (INIS)

    Urli, N.; Pevec, D.; Coffou, E.; Petrovic, B.

    1980-01-01

    The most important methods in the LWR in-core fuel management are reviewed. A modular approach and optimization by use of infinite multiplication factor and power form-factor are favoured. A computer program for rotation of fuel assemblies at reloads has been developed which improves further fuel economy and reliability of nuclear power plants. The program has been tested on the PWR core and showed to decrease the power form-factors and flatten the radial power distribution. (author)

  7. Achilles tests finally nail PWR fuel clad ballooning fears

    International Nuclear Information System (INIS)

    Dore, P.; McMinn, K.

    1992-01-01

    A conclusive series of experiments carried out by AEA Reactor Services at its Achilles rig in the UK has finally allayed fears that fuel clad ballooning is a major safety problem for Sizewell B, Britain's first Pressurized Water Reactor. The experiments are described in this article. (author)

  8. 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

  9. 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.

  10. 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

  11. 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)

  12. 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

  13. 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

  14. 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)

  15. 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.

  16. 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

  17. 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

  18. 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)

  19. SODEXPERT: help to PWR plant management to prevent secondary circuit corrosion

    International Nuclear Information System (INIS)

    Eon-Duval, P.; Fiquet, J.M.; Langlet, J.P.

    1996-01-01

    Since about 10 years, problems of secondary circuit corrosion have raised for PWR plant management. The watch staff can't be asked the physicochemical knowledge requested for a proper interpretation of the various probes outputs. So an expert-system has been performed to help the identification of dangerous situation from a corrosion point of view, and immediately start the PWR managing action. This software has been successfully tested and validated. (D.L.). 5 figs., 4 photos

  20. 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.

  1. 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

  2. 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.

  3. 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

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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)

  9. 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.

  10. 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

  11. 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

  12. Burnup Credit of French PWR-MOx fuels: methodology and associated conservatisms with the JEFF-3.1.1 evaluation

    International Nuclear Information System (INIS)

    Chambon, A.

    2013-01-01

    Considering spent fuel management (storage, transport and reprocessing), the approach using 'fresh fuel assumption' in criticality-safety studies results in a significant conservatism in the calculated value of the system reactivity. The concept of Burnup Credit (BUC) consists in considering the reduction of the spent fuel reactivity due to its burnup. A careful BUC methodology, developed by CEA in association with AREVA-NC was recently validated and written up for PWR-UOx fuels. However, 22 of 58 French reactors use MOx fuel, so more and more irradiated MOx fuels have to be stored and transported. As a result, why industrial partners are interested in this concept is because taking into account this BUC concept would enable for example a load increase in several fuel cycle devices. Recent publications and discussions within the French BUC Working Group highlight the current interest of the BUC concept in PWR-MOx spent fuel industrial applications. In this case of PWR-MOx fuel, studies show in particular that the 15 FPs selected thanks to their properties (absorbing, stable, non-gaseous) are responsible for more than a half of the total reactivity credit and 80% of the FPs credit. That is why, in order to get a conservative and physically realistic value of the application k eff and meet the Upper Safety Limit constraint, calculation biases on these 15 FPs inventory and individual reactivity worth should be considered in a criticality-safety approach. In this context, thanks to an exhaustive literature study, PWR-MOx fuels particularities have been identified and by following a rigorous approach, a validated and physically representative BUC methodology, adapted to this type of fuel has been proposed, allowing to take fission products into account and to determine the biases related to considered isotopes inventory and to reactivity worth. This approach consists of the following studies: - isotopic correction factors determination to guarantee the criticality

  13. 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)

  14. 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.

  15. 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)

  16. 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)

  17. 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%

  18. CLUB, Cell Calculation PF Candu PWR Fuel Clusters

    International Nuclear Information System (INIS)

    Krishnani, P.D.

    1985-01-01

    1 - Description of problem or function: CLUB is an integral transport theory code to calculate fluxes, reaction rates and few-group condensed Cross sections for cylindricalized PHWR lattice cells. For a specified buckling, it computes k eff using few-group diffusion theory in fundamental mode. There is also an option to calculate these quantities as a function of burnup. 2 - Method of solution: There are basically two options for solving the integral equation. In the first option, the integral transport equation is solved by the combination of the small scale Pij method and the large scale interface current technique. At each region interface, the angular flux is expanded separately in the incoming and outgoing direction. Up to three terms can be considered in this expansion. In the second option, the complete Pij method is used for the cylindricalized lattice cell. The calculations are performed in 27 groups for which the Cross sections are derived from the 69-group WIMS library by condensing them into 27 groups by using a typical spectrum of PHWRs. The first order differential burnup equations can be solved by either the trapezoidal rule or the Runge-Kutta method. 3 - Restrictions on the complexity of the problem: The program considers the same number of zones in each ring. Furthermore, the fuel pin in each ring should be of the same type

  19. 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)

  20. 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.)

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. 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

  6. The application of neural networks for optimization of the configuration of fuel assemblies in PWR reactors

    International Nuclear Information System (INIS)

    Sadighi, M.; Setayeshi, S.; Salehi, A.A.

    2002-01-01

    This paper presents a new method to solve the problem of finding the best configuration of fuel assemblies in a PWR (Pressurized Water Reactor) core. Finding an optimum solution requires a huge amount of calculations in classical methods. It has been shown that the application of continuous Hop field neural network accompanied by the Simulated Annealing method to this problem not only reduces the volume of the calculations, but also guarantees finding the best solution. In this study flattening of neutron flux inside the reactor core of Brusher NPP is considered as an objective function. The result shows the optimum core configuration which is in agreement with the pattern proposed by the designer

  7. 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)

  8. 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)

  9. 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.

  10. 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

  11. Identification and evaluation of PWR in-vessel severe accident management strategies

    International Nuclear Information System (INIS)

    Dukelow, J.S.; Harrison, D.G.; Morgenstern, M.

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents

  12. In-core fuel management: New challenges

    International Nuclear Information System (INIS)

    Kolmayer, A.; Vallee, A.; Mondot, J.

    1992-01-01

    Experience accumulated by pressurized water reactor (PWR) utilities allows them to improve their strategies in the use of eventual margins to core design limits. They are used for nuclear steam supply system (NSSS) power upgrading, to improve operating margins, or to adapt fuel management to specific objectives. As a result, in-core fuel management strategies have become very diverse: UO 2 or mixed-oxide loading, out-in or in-out fuel loading patterns, extended or annual cycle lengths with margins on design limits such as moderator temperature coefficients, boron concentrations, or peaking factors. Perspectives also appear concerning use of existing plutonium stocks or actinide incineration. Burnable poisons are most often needed to satisfactorily achieve these goals. Among them, gadolinia are now largely used, owing to their excellent performance. More than 24 Framatome first cores and reloads, representing more than 3000 gadolinia-bearing rods, have been irradiated since 1983

  13. Fuel management and economics

    Energy Technology Data Exchange (ETDEWEB)

    Vendryes, G

    1972-11-01

    From international conference on nuclear solutions to world energy problems; Washington, District of Columbia, USA (12 Nov The low cost of the fuel cycle is the most attractive feature of the fast neutron breeder reactor. In order to achieve it a good fuel management is essential, with well balanced fixed investment and renewal fuel costs. In addition the designer can optimize the power station as a whole (fuel cycle and thermal characteristics). (auth)

  14. 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)

  15. 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)

  16. 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)

  17. Fuel management and core design code systems for pressurized water reactor neutronic calculations

    International Nuclear Information System (INIS)

    Ahnert, C.; Arayones, J.M.

    1985-01-01

    A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

  18. 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)

  19. 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.

  20. 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.

  1. French nuclear plants PWR vessel integrity assessment and life management

    Energy Technology Data Exchange (ETDEWEB)

    Bezdikian, G. [Electricite de France (EDF), Div. Production Nucleaire, 93 - Saint-Denis (France); Quinot, P. [FRAMATOME, Dept. Bloc Reacteur et Boucles Primaires, 92 - Paris-La-Defence (France); Faidy, C.; Churier-Bossennec, H. [Electricite de France (EDF), Div. Ingenierie et Service, 69 - Villeurbanne (France)

    2001-07-01

    The Reactor Pressure Vessel life management of 56 PWR 3 loop and 4 loop reactors units was engaged by the French Utility EDF (Electricite de France) a few years ago and is yet on going on. This paper will present the work carried out within the framework of justifying why the 34 three loop reactor vessels will remain acceptable for operation for a lifetime of at least 40-years. A summary of the measures will be given. An overall review of actions will be presented describing the French approach, using important existing databases, including studies related to irradiation surveillance monitoring program and end of life fluence assessment. The last results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions) until the end of lifetime, postulating circumferential an radial kinds of flaw located in the stainless steel cladding or shallow sub-cladding area. The results of structural integrity analyses beginning with elastic computations and completed with three-dimensional finite element elastic plastic computations for envelope cases, are compared with code criteria for operating plants. The objective is to evaluate the margins on different parameters as RTNDT (Reference Nil Ductility Transition Temperature), toughness or crack size, to justify the global fitness for service of all these Reactor Pressure Vessels. The paper introduces EDF's maintenance strategy, related to integrity assessment, for those nuclear power plants under operation, based on NDE in-service inspection of the first thirty millimeters in the thickness of the wall and major surveillance programs of the vessels. (author)

  2. French nuclear plants PWR vessel integrity assessment and life management

    International Nuclear Information System (INIS)

    Bezdikian, G.; Quinot, P.; Faidy, C.; Churier-Bossennec, H.

    2001-01-01

    The Reactor Pressure Vessel life management of 56 PWR 3 loop and 4 loop reactors units was engaged by the French Utility EDF (Electricite de France) a few years ago and is yet on going on. This paper will present the work carried out within the framework of justifying why the 34 three loop reactor vessels will remain acceptable for operation for a lifetime of at least 40-years. A summary of the measures will be given. An overall review of actions will be presented describing the French approach, using important existing databases, including studies related to irradiation surveillance monitoring program and end of life fluence assessment. The last results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions) until the end of lifetime, postulating circumferential an radial kinds of flaw located in the stainless steel cladding or shallow sub-cladding area. The results of structural integrity analyses beginning with elastic computations and completed with three-dimensional finite element elastic plastic computations for envelope cases, are compared with code criteria for operating plants. The objective is to evaluate the margins on different parameters as RTNDT (Reference Nil Ductility Transition Temperature), toughness or crack size, to justify the global fitness for service of all these Reactor Pressure Vessels. The paper introduces EDF's maintenance strategy, related to integrity assessment, for those nuclear power plants under operation, based on NDE in-service inspection of the first thirty millimeters in the thickness of the wall and major surveillance programs of the vessels. (author)

  3. 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)

  4. 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)

  5. 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.)

  6. 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).

  7. 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.

  8. 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

  9. 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

  10. 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

  11. Utilization of ''CONTACT'' experiments to improve the fission gas release knowledge in PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Charles, M; Abassin, J J; Bruet, M; Baron, D; Melin, P

    1983-03-01

    The CONTACT experiments, which were carried out by the French CEA, within the framework of a CEA-FRAMATOME collaboration agreement, bear on the behaviour of in-pile irradiated PWR fuel rods. We will focus here upon their results dealing with fission gas release. The experimental device is briefly described, then the following results are given: the kinetics of stable fission gas release for various linear ratings; the instantaneous fractional release rates of radioactive gases versus their decay constant in the range 1.5 10/sup -6/-3.6 10/sup -3/s/sup -1/, for various burnups, as also the influence of fuel temperature. Moreover, the influence of the nature and the pressure of the filling gas upon the release is presented for various linear ratings. The experimental results are discussed and analysed with the purpose to model various physical phenomena involved in the release (low-temperature mechanisms, diffusion).

  12. 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

  13. 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

  14. Utilization of ''CONTACT'' experiments to improve the fission gas release knowledge in PWR fuel rods

    International Nuclear Information System (INIS)

    Charles, M.; Abassin, J.J.; Bruet, M.

    1983-01-01

    The CONTACT experiments, which were carried out by the French CEA, within the framework of a CEA-FRAMATOME collaboration agreement, bear on the behaviour of in-pile irradiated PWR fuel rods. We will focus here upon their results dealing with fission gas release. The experimental device is briefly described, then the following results are given: the kinetics of stable fission gas release for various linear ratings; the instantaneous fractional release rates of radioactive gases versus their decay constant in the range 1.5 10 -6 -3.6 10 -3 s -1 , for various burnups, as also the influence of fuel temperature. Moreover, the influence of the nature and the pressure of the filling gas upon the release is presented for various linear ratings. The experimental results are discussed and analysed with the purpose to model various physical phenomena involved in the release (low-temperature mechanisms, diffusion)

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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

  1. Development of advanced spent fuel management process

    International Nuclear Information System (INIS)

    Ro, Seung Gy; Shin, Y. J.; Do, J. B.; You, G. S.; Seo, J. S.; Lee, H. G.

    1998-03-01

    This study is to develop an advanced spent fuel management process for countries which have not yet decided a back-end nuclear fuel cycle policy. The aims of this process development based on the pyroreduction technology of PWR spent fuels with molten lithium, are to reduce the storage volume by a quarter and to reduce the storage cooling load in half by the preferential removal of highly radioactive decay-heat elements such as Cs-137 and Sr-90 only. From the experimental results which confirm the feasibility of metallization technology, it is concluded that there are no problems in aspects of reaction kinetics and equilibrium. However, the operating performance test of each equipment on an engineering scale still remain and will be conducted in 1999. (author). 21 refs., 45 tabs., 119 figs

  2. 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

  3. Study on entry criteria for severe accident management during hot leg LBLOCAs in a PWR

    International Nuclear Information System (INIS)

    Zhang, Longfei; Zhang, Dafa; Wang, Shaoming

    2007-01-01

    The risk of Large Break Loss of Coolant Accidents (LBLOCA) has been considered an important safety issue since the beginning of the nuclear power industry. The rapid depressurization occurs in the primary coolant circuit when a large break appears in a Pressurized Water Reactors (PWR).Then the coolant temperature reaches saturation at a very low pressure. The core outlet fluid temperatures maybe not reliable indicators of the core damage states at a such lower pressure. The problem is how to decide the time for water injection in the SAM (Severe Accident Management). An alternative entry criterion is the fluid temperature just above the hot channel in which the fluid temperature showed maximum among all the channels. For that reason, a systematic study of entry criterion of SAM for different hot leg break sizes in a 3-loop PWR has been started using the detailed system thermal hydraulic and severe accident analysis code package, RELAP/SCDAPSIM. Best estimate calculations of the large break LOCA of 15 cm, 20 cm and 25 cm without accident managements and in the case of high-pressure safety injection as the accident management were performed in this paper. The analysis results showed that the core exit temperatures are not reliable indicators of the peak core temperatures and core damage states once peak core temperatures reach 1500 K, and the proposed entry criteria for SAM at the time when the core outlet temperature reaches 900 K is not effective to prevent core melt. Then other analyses were performed with a parameter of fluid temperature just above the hot channel. The latter analysis showed that earlier water injection when the fluid temperature just above the hot channel reaches 900 K is effective to prevent further core melt. Since fuel surface and hot channel have spatial distribution and depend on a period of cycle operation, a series of thermocouples are required to install just above the fuel assembly. The maximum exit temperature of 900 K that captured by

  4. On the problem of in-core fuel management in power reactors

    International Nuclear Information System (INIS)

    Marinkovic, N.; Matausek, M.V.

    1985-01-01

    Within the scope of in-core fuel management including refuelling schedule and reactivity control it is indispensable to define nuclear fuel worth, optimal depletion of the spent fuel assemblies as well as isotopic composition of the spent fuel. This paper shows the computed values of the mentioned parameters in case of different reactor types, PWR, WWER, HWR and BWR of 1000 MWe as well as the intensity of radiation of the spent fuel 3 and 1 years after fission.(author)

  5. Fuel cycle management

    International Nuclear Information System (INIS)

    Herbin, H.C.

    1977-01-01

    The fuel cycle management is more and more dependent on the management of the generation means among the power plants tied to the grid. This is due mainly because of the importance taken by the nuclear power plants within the power system. The main task of the fuel cycle management is to define the refuelling pattern of the new and irradiated fuel assemblies to load in the core as a function of: 1) the differences which exist between the actual conditions of the core and what was expected for the present cycle, 2) the operating constraints and the reactor availability, 3) the technical requirements in safety and the technological limits of the fuel, 4) the economics. Three levels of fuel cycle management can be considered: 1) a long term management: determination of enrichments and expected cycle lengths, 2) a mid term management whose aim corresponds to the evaluation of the batch to load within the core as a function of both: the next cycle length to achieve and the integrated power history of all the cycles up to the present one, 3) a short term management which deals with the updating of the loaded fuel utilisations to take into account the operation perturbations, or with the alteration of the loading pattern of the next batch to respect unexpected conditions. (orig.) [de

  6. 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)

  7. 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

  8. Spent fuel management

    International Nuclear Information System (INIS)

    2005-01-01

    The production of nuclear electricity results in the generation of spent fuel that requires safe, secure and efficient management. Appropriate management of the resulting spent fuel is a key issue for the steady and sustainable growth of nuclear energy. Currently about 10,000 tonnes heavy metal (HM) of spent fuel are unloaded every year from nuclear power reactors worldwide, of which 8,500 t HM need to be stored (after accounting for reprocessed fuel). This is the largest continuous source of civil radioactive material generated, and needs to be managed appropriately. Member States have referred to storage periods of 100 years and even beyond, and as storage quantities and durations extend, new challenges arise in the institutional as well as in the technical area. The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs

  9. Fuel assembly loads during a hypothetical blowdown event in a PWR

    International Nuclear Information System (INIS)

    Stabel, J.; Bosanyi, B.; Kim, J.D.

    1991-01-01

    As a consequence of a hypothetical sudden break of the main coolant pipe of a PWR, RPV-internals and fuel assemblies (FA's) are undergoing horizontal and vertical motions. FA's may impact against each other, against core shroud or against lower core support. The corresponding impact loads must be absorbed by the FA spacer grids and guide thimbles. In this paper FA-loads are calculated with and without consideration of Fluid-Structure-Interaction (FSI) effects for assumed different break sizes of the main coolant pipe. The analysis has been performed for a hypothetical cold leg break of a typical SIEMENS-4 loop plant. For this purpose the codes DAPSY/DAISY (GRS, Germany) were coupled with the structural code KWUSTOSS (SIEMENS). It is shown that the FA loads obtained in calculations with consideration of FSI effects are by a factor of 2-4 lower than those obtained in the corresponding calculations without consideration of FSI. (author)

  10. 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)

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

    International Nuclear Information System (INIS)

    Clavier, B.

    1995-01-01

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

  12. 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

  13. Status of spent fuel management in the United Kingdom

    International Nuclear Information System (INIS)

    Dodds, R.

    1996-01-01

    Nuclear generating capacity in the UK is static with no units currently under construction following the completion of the Sizewell B PWR. The Government's reviews of nuclear energy policy and radioactive waste management policy have been published following a public consultation procedure, largely with an endorsement of current policies. Nuclear Electric plc (NE) and Scottish Nuclear Limited's (SNL) AGR and PWR stations are to be privatised as two subsidiaries of a holding company, and it is planned that the Magnox stations and their liabilities will be kept in the public sector, initially in a stand alone company but ultimately integrated with BNFL. Prompt reprocessing of all Magnox fuel will continue. NE and SNL have signed contracts for extensive reprocessing of AGR fuel. In addition, SNL has agreed contractual arrangements with BNFL for long term storage of its remaining overlife arisings of AGR fuel and has therefore on commercial ground opted not to dry store their fuel at the reactor site. NE have not yet made a decision on the fate of their AGR fuel not covered by existing reprocessing contracts. No option selection has taken place for PWR fuel. Following the closure of the Dounreay PFR and the withdrawal from the EFR project, the option of recycle of plutonium in fast reactors has been suspended. (author)

  14. Vibration characteristics of a PWR fuel rod supported by optimized H type spacer grids

    International Nuclear Information System (INIS)

    Choi, M. H.; Kang, H. S.; Yoon, K. H.; Kim, H. K.; Song, K. N.

    2002-01-01

    The spacer grids are one of the main structural components in the fuel assembly, which supports and protects the fuel rods from the external loads by seismic and coolant flow. In this study, a modal test and a FE vibration analysis using ABAQUS are performed on a PWR dummy fuel rod of 3.847 m which is continuously supported by eight Optimized H type spacer grids. The experimental results agree with previous works that the natural frequencies decrease, while the amplitudes increase, with the increase of the excitation force. The force levels showing the maximum displacement of 0.2 mm are in the range from 0.2 N to 0.3 N, and at the same force range the fundamental frequencies are measured around 42.0 Hz, at which the relatively big displacements are observed at the 7th span. The results from the modal tests and the FE analyses are compared by both Modal Assurance Criteria (MAC) values and mode shapes. The MAC values at 2nd, 4th, and 7th mode are below 50%. It is believed that the reason of the low MACs at those modes is that the vibration amplitudes of the modes are more distorted by the excitation force than those of the other modes

  15. 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)

  16. High mechanical performance of Areva upgraded fuel assemblies for PWR in USA

    International Nuclear Information System (INIS)

    Gottuso, Dennis; Canat, Jean-Noel; Mollard, Pierre

    2007-01-01

    The merger of the product portfolios of the former Siemens and Framatome fuel businesses gave rise to a new family of PWR products which combine the best features of the different technologies to enhance the main performance of each of the existing products. In this way, the technology of each of the three main fuel assembly types usually delivered by AREVA NP, namely Mark-BW TM , HTP TM and AFA 3G TM has been enriched by one or several components from the others which contributes to improve their robustness and to enhance their performance. The combined experience of AREVA's products shows that the ROBUST FUELGUARD TM , the HMP TM end grid, the MONOBLOC TM guide tube, a welded structure, M5 R material for every zirconium component and an upper QUICK-DISCONNECT TM are key features for boosting fuel assembly robustness. The ROBUST FUELGUARD benefits from a broad experience demonstrating its high efficiency in stopping debris. In addition, its mechanical strength has been enhanced and the proven blade design homogenizes the downstream flow distribution to strongly reduce excitation of fuel rods. The resistance to rod-to-grid fretting resistance of AREVA's new products is completed by the use of a lower HMP grid with 8 lines of contact to insure low wear. The Monobloc guide tube with a diameter maximized to strengthen the fuel assembly stiffness, excludes through its uniform outer geometry any local condition which could weaken guide tube straightness. The application of a welded cage to all fuel assemblies of the new family of products in combination with stiffer guide tubes and optimized hold-down assures each fuel assembly enhanced resistance to distortion. The combination of these features has been widely demonstrated as an effective method to reduce the risk of incomplete RCCA insertion and significantly reduce assembly distortion. Thanks to its enhanced performance, M5 alloy insures that all fuel assemblies in the family maintain their performance in all

  17. Fission gas induced fuel swelling in low and medium burnup fuel during high temperature transients. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Vinjamuri, K.

    1980-01-01

    The behavior of light water reactor fuel elements under postulated accident conditions is being studied by the EG and G Idaho, Inc., Thermal Fuels Behavior Program for the Nuclear Regulatory Commission. As a part of this program, unirradiated and previously irradiated, pressurized-water-reactor type fuel rods were tested under power-cooling-mismatch (PCM) conditions in the Power Burst Facility (PBF). During these integral in-reactor experiments, film boiling was produced on the fuel rods which created high fuel and cladding temperatures. Fuel rod diameters increased in the film boiling region to a greater extent for irradiated rods than for unirradiated rods. The purpose of the study was to investigate and assess the fuel swelling which caused the fuel rod diameter increases and to evaluate the ability of an analytical code, the Gas Release and Swelling Subroutine - Steady-State and Transient (GRASS-SST), to predict the results.

  18. 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

  19. 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

  20. 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)

  1. Zircaloy PWR fuel cladding deformation tests under mainly convective cooling conditions

    International Nuclear Information System (INIS)

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

    1980-01-01

    In a loss-of-coolant accident the temperature of the cladding of the fuel rods may rise to levels (650-810 0 C) where the ductility of Zircaloy is high (approximately 80%). The net outward pressure which will obtain if the coolant pressure falls to a small fraction of its normal working value produces stresses in the cladding which can result in large strain through secondary creep. An earlier study of the deformation of specimens of PWR Zircaloy cladding tubing 450 mm long under internal pressure had shown that strains of over 50% could be produced over considerable lengths (greater than twenty tube diameters). Extended deformation of this sort might be unacceptable if it occurred in a fuel element. The previous tests had been carried out under conditions of uniform radiative heat loss, and the work reported here extends the study to conditions of mainly convective heat loss believed to be more representative of a fuel element following a loss of coolant. Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 845 0 C in flowing steam at atmospheric pressure. Internal test pressures were in the range 2.9-11.0 MPa (400-1600 1b/in 2 ). Maximum strains were observed of the same magnitude as those seen in the previous tests, but the shape of the deformation differed; in these tests the deformation progressively increased in the direction of the steam flow. These results are compared with those from multi-rod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behaviour of fuel elements in a loss-of-coolant accident are outlined. (author)

  2. TWODEE-2/MOD3, 2-D Time-Dependent Fuel Elements Thermal Analysis after PWR LOCA

    International Nuclear Information System (INIS)

    Lauben, G. N.

    2001-01-01

    1 - Description of problem or function: WREM-TOODEE2 is a two- dimensional, time-dependent, fuel-element thermal analysis program. Its primary purpose is to evaluate fuel-element thermal response during post-LOCA refill and reflood in a pressurized water reactor (PWR). 2 - Method of solution: TOODEE2 calculations are carried out in a two-dimensional mesh region defined in slab or cylindrical geometry by orthogonal grid lines. Coordinates which form order pairs are labeled x-y in slab geometry, and those in cylindrical geometry are labeled r-z for the axisymmetric case and r-theta for the polar case. Conduction and radiation are the only heat transfer mechanisms assumed within the boundaries of the mesh region. Convective and boiling heat transfer mechanisms are assumed at the boundaries. The program numerically solves the two-dimensional, time-dependent, heat conduction equation within the mesh region. 3 - Restrictions on the complexity of the problem: WREM-TOODEE2 considers only axisymmetric geometry although the equations for slab and polar geometry are included in the program

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

    International Nuclear Information System (INIS)

    Foad, Basma; Takeda, Toshikazu

    2015-01-01

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

  4. Experience in the use of low concentration gadolinia as a PWR fuel burnable absorber

    International Nuclear Information System (INIS)

    Mildrum, C.M.; Segovia, M.A.

    2001-01-01

    A description is provided of the low concentration gad design being used in the Spanish 3-loop 17 x 17 fueled PWR's. This design uses a relatively small number of high concentration gadolinia fuel rods (6 and 8 w/o Gd2O3) with a large number of low concentration gad rods (2 w/o Gd2O3). The 2 w/o gad rods substitute, in part, the high concentration gad rods, thereby helping reduce the end of cycle reactivity penalty from the residual absorption in the gadolinium. The low concentration gad design is advantageous for long cycles (more than 18 months) and plant up-rating scenarios in that the soluble boron concentration increases that would otherwise result for these situations are avoided. These boron concentration increases could have potentially adverse effects on the plant, since the moderator temperature coefficient (MTC) is made less negative, the effectiveness of the boron shutdown safety systems is reduced, and the safety margins are eroded for some accidents, such as for boron dilution events. This paper also reviews the APA nuclear design code system performance for the low concentration gad design. (author)

  5. 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

  6. AREVA modeling and predictive capacities to support PWR fuel assembly upgrading

    International Nuclear Information System (INIS)

    Canat, J. N.; Mollard, P.; Gentet, G.; Uyeda, G.

    2008-01-01

    optimization. It explains how they have supported AREVA in performing a wide ranging review of the various components resulting from the technologies acquired after the merging of FRAMATOME with Siemens Nuclear activities (now resulting in AREVA NP): AFA 3G TM , HTP TM and Mark B TM . Today, AREVA offers to each reactor worldwide the best suited fuel assembly, fully assessed and featuring upgraded components issued from well proven technologies, such as AGORA fuel assemblies or other very advanced versions of fuel assemblies. In all areas of fuel design, computational codes, methods and modeling have improved. The paper details the main progress achieved in a variety of fields. Special attention will be paid to improvements performed in the thermal hydraulics area where the Computational Fluid Dynamics STAR CD tool has been used in several Fuel Assembly design and development applications. This progress was made possible by recently available computer hardware, allowing cost effective and easy access of parallel computing in cluster arrays. Specific applications such as the optimization of the vane shape of PWR grids for higher DNB performance, the evaluation of pressure loss coefficients for complex components such as the anti debris bottom nozzle, computation of the flow velocity distribution and consequent fuel rod vibration regime will be detailed. As for fuel rod design, 3 D modeling of the PCI phenomenon provides a basis for improvements of industrial 1 D predictive tools with a better level of accuracy in discriminating hardware solutions. The developments of the next product generations are reaping the benefits of all these improvements, paying off in optimized designs and upgraded performance and robustness levels, for the final benefit of the utilities, especially with the prospect of fuel supply to the forthcoming Generation III reactors

  7. 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)

  8. 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)

  9. Research on nuclear energy in the fields of fuel cycle, PWR reactors and LMFBR reactors

    International Nuclear Information System (INIS)

    Barre, B.; Camarcat, N.

    1995-01-01

    In this article we present the CEA research programs to improve the safety of the next generation of reactors, to manage the Plutonium and the wastes of the fuel cycle end and to ameliorate the competitiveness. 6 refs

  10. 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)

  11. Potential impacts of crud deposits on fuel rod behaviour on high powered PWR fuel rods

    International Nuclear Information System (INIS)

    Wilson, W.; Comstock, R.J.

    1999-01-01

    Fuel assemblies operating with significant sub-cooled boiling are subject to deposition of surface deposits commonly referred to as crud. This crud can potentially cause concentration of chemical species within the deposits which can be detrimental to cladding performance in PWRs. In addition, these deposits on the surface of the cladding can result in power anomalies and erroneous reporting of fuel rod oxide thickness which can substantially hamper corrosion and core performance modeling efforts. Data is presented which illustrates the importance of accounting for the presence of crud on fuel cladding surfaces. Several methods used to correct for this phenomenon when collecting and analyzing zirconium alloy field oxide thickness measurements are described. Various observations related to crud characteristics and its impact on fuel rod performance are also addressed. (author)

  12. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    International Nuclear Information System (INIS)

    Bi, G.; Liu, C.; Si, S.

    2012-01-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade 233 U-Thorium (U 3 ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade 233 U extracted from burnt PuThOX fuel was used to fabrication of U 3 ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U 3 ThOX mixed core, the well designed U 3 ThOX FAs with 1.94 w/o fissile uranium (mainly 233 U) were located on the periphery of core as a blanket region. U 3 ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U 3 ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U 3 ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U 3 ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U 3 ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared

  13. Experience in the use of low concentration gadolinia as a PWR fuel burnable absorber

    International Nuclear Information System (INIS)

    Mildrum, C.M.; Segovia, M.A.

    2001-01-01

    A description is provided of the low concentration gad design being used in the Spanish 3-loop 17 x 17 fueled PWR's. This design uses a relatively small number of high concentration gadolinia fuel rods (6 and 8 w/o Gd 2 O 3 ) with a large number of low concentration gad rods (2 w/o Gd 2 O 3 ). The 2 w/o gad rods substitute, in part, the high concentration gad rods, thereby helping reduce the end of cycle reactivity penalty from the residual absorption in the gadolinium. The low concentration gad design is advantageous for long cycles (18+ months) and plant up-rating scenarios in that the soluble boron concentration increases that would otherwise result for these situations are avoided. These boron concentration increases could have potentially adverse effects on the plant, since the moderator temperature coefficient (MTC) is made less negative, the effectiveness of the boron shutdown safety systems is reduced, and the safety margins are eroded for some accidents, such as for boron dilution events. These increases in the boron concentration would also require the plant to operate at higher lithium (Li) concentrations in the coolant in order to maintain the pH level at the desired value. Operation at the higher Li concentrations is undesirable because of the concerns over the potential impact on the fuel assembly material performance (e.g., crud and corrosion). This paper also reviews the APA (Alpha/Phoenix-P/ANC) nuclear design code system performance for the low concentration gad design. The design system performance for the reload cores that have or are employing this design has been completely satisfactory. The performance and accuracy of the nuclear design methodology is found to be as good for this design as for the reload cores that use exclusively high gad concentrations, or those that use WABA's - the discrete burnable absorber (BA) used prior to its substitution for gadolinium. (authors)

  14. Heat transfer in a spent fuel pool concept containing PWR, Hybrid ADS-Fission, and VHTR spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Fernando P.; Cardoso, Fabiano; Salomé, Jean A.D.; Velasquez, Carlos E.; Pereira, Claubia, E-mail: fernandopereirabh@gmail.com, E-mail: fabinuclear@yahoo.com.br, E-mail: jadsalome@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Thermal evaluation under wet storage conditions of spent fuels (SF) of the types UO{sub 2} discharged from Pressurized Water Reactor (PWR) and Very High-temperature Reactor (VHTR), and (Th,TRU)O{sub 2} from Accelerator-Driven Subcritical Reactor System (ADS) and VHTR are presented. The analyzes are in the absence of an external cooling system of the pool, and the goal is to compare the water boiling time of the pool storing these different types of SF, at time t=0 year after reactor discharge. Two techniques were implemented. In the first one, all the materials of the fuel elements are considered. In the second, the SF is treated as holes inside the pool, assuming the heat transfer directly from the SF to the water. Results from first technique show that the boiling time (T{sub b}) ranged from 23 minutes for (Th,TRU)O{sub 2} from VHTR to 3 hours for UO{sub 2} from VHTR, while for the second technique, T{sub b} ranged from 10 minutes for (Th,TRU)O{sub 2} from VHTR to 2.7 hours for UO{sub 2} from VHTR. The discrepancies between Tb from both techniques reveal that the pathways considered for the heat transfer are crucial to the results. The thermal studies used the module CFX of the ANSYS Workbench 16.2 - student version. (author)

  15. 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)

  16. 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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  18. Pre-conceptual design of a spent PWR fuel disposal container

    International Nuclear Information System (INIS)

    Choi, Jong Won; Cho, Dong Keun; Lee, Yang; Choi, Heui Joo; Lee, Jong Youl

    2005-01-01

    In this paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid and bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert. the Maximum Von Mises stress from the 102 cm container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by ∼20 tons

  19. Status analysis for the confinement monitoring technology of PWR spent nuclear fuel dry storage system

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-15

    Leading national R and D project to design a PWR spent nuclear fuel interim dry storage system that has been under development since mid-2009, which consists of a dual purpose metal cask and concrete storage cask. To ensure the safe operation of dry storage systems in foreign countries, major confinement monitoring techniques currently consist of pressure and temperature measurement. In the case of a dual purpose metal cask, a pressure sensor is installed in the interspace of bolted double lid(primary and secondary lid) in order to measure pressure. A concrete storage cask is a canister based system made of double/redundant welded lid to ensure confinement integrity. For this reason, confinement monitoring method is real time temperature measurement by thermocouple placed in the air flow(air intake and exit) of the concrete structure(over pack and module). The use of various monitoring technologies and operating experiences for the interim dry storage system over the last decades in foreign countries were analyzed. On the basis of the analysis above, development of the confinement monitoring technology that can be used optimally in our system will be available in the near future.

  20. Determination of uncertainties of PWR spent fuel radionuclide inventory based on real operational history data

    International Nuclear Information System (INIS)

    Fast, Ivan; Bosbach, Dirk; Aksyutina, Yuliya; Tietze-Jaensch, Holger

    2015-01-01

    A requisite for the official approval of the safe final disposal of SNF is a comprehensive specification and declaration of the nuclear inventory in SNF by the waste supplier. In the verification process both the values of the radionuclide (RN) activities and their uncertainties are required. Burn-up (BU) calculations based on typical and generic reactor operational parameters do not encompass any possible uncertainties observed in real reactor operations. At the same time, the details of the irradiation history are often not well known, which complicates the assessment of declared RN inventories. Here, we have compiled a set of burnup calculations accounting for the operational history of 339 published or anonymized real PWR fuel assemblies (FA). These histories were used as a basis for a 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's). Hence, we can calculate the realistic variation or spectrum of RN inventories. SCALE 6.1 has been employed for the burn-up calculations. The results have been validated using experimental data from the online database - SFCOMPO-1 and -2. (authors)

  1. Development of built-in debris-filter bottom nozzle for PWR fuel assemblies

    International Nuclear Information System (INIS)

    Juntaro Shimizu; Kazuki Monaka; Masaji Mori; Kazuo Ikeda

    2005-01-01

    Mitsubishi Heavy Industries, Ltd. (MHI) has worked to improve the capability of anti debris bottom nozzle for a PWR fuel assembly. The Current debris filter bottom nozzle (DFBN) having 4mm diameter flow holes can capture the larger size of debris than the flow hole inner diameter. MHI has completed the development of the built-in debris filter bottom nozzle, which is the new idea of the debris-filter for high burnup (55GWd/t assembly average burnup). Built-in debris filter bottom nozzle consists of the blades and nozzle body. The blades made from inconel strip are embedded and welded on the grooved top surface of the bottom nozzle adapter plate. A flow hole is divided by the blade and the trap size of the debris is reduced. Because the blades block the coolant flow, it was anticipated to increase the pressure loss of the nozzle, however, adjusting the relation between blade and taper shape of the flow hole, the pressure loss has been successfully maintained the satisfactory level. Grooves are cut on the nozzle plate; nevertheless, the additional skirts on the four sides of the nozzle compensate the structural strength. (authors)

  2. Determination of uncertainties of PWR spent fuel radionuclide inventory based on real operational history data

    Energy Technology Data Exchange (ETDEWEB)

    Fast, Ivan; Bosbach, Dirk [Institute of Energy- and Climate Research, Nuclear Waste Management and Reactor Safety Research, IEK-6, Forschungszentrum, Julich GmbH, (Germany); Aksyutina, Yuliya; Tietze-Jaensch, Holger [German Product Control Office for Radioactive Waste (PKS), Institute of Energy- and Climate Research, Nuclear Waste Management and Reactor Safety Research, IEK-6, Forschungszentrum Julich GmbH, (Germany)

    2015-07-01

    A requisite for the official approval of the safe final disposal of SNF is a comprehensive specification and declaration of the nuclear inventory in SNF by the waste supplier. In the verification process both the values of the radionuclide (RN) activities and their uncertainties are required. Burn-up (BU) calculations based on typical and generic reactor operational parameters do not encompass any possible uncertainties observed in real reactor operations. At the same time, the details of the irradiation history are often not well known, which complicates the assessment of declared RN inventories. Here, we have compiled a set of burnup calculations accounting for the operational history of 339 published or anonymized real PWR fuel assemblies (FA). These histories were used as a basis for a 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's). Hence, we can calculate the realistic variation or spectrum of RN inventories. SCALE 6.1 has been employed for the burn-up calculations. The results have been validated using experimental data from the online database - SFCOMPO-1 and -2. (authors)

  3. ORCOST-2, PWR, BWR, HTGR, Fossil Fuel Power Plant Cost and Economics

    International Nuclear Information System (INIS)

    Fuller, L.C.; Myers, M.L.

    1975-01-01

    1 - Description of problem or function: ORCOST2 estimates the cost of electrical energy production from single-unit steam-electric power plants. Capital costs and operating and maintenance costs are calculated using base cost models which are included in the program for each of the following types of plants: PWR, BWR, HTGR, coal, oil, and gas. The user may select one of several input/output options for calculation of capital cost, operating and maintenance cost, levelized energy costs, fixed charge rate, annual cash flows, cumulative cash flows, and cumulative discounted cash flows. Options include the input of capital cost and/or fixed charge rate to override the normal calculations. Transmission and distribution costs are not included. Fuel costs must be input by the user. 2 - Method of solution: The code follows the guidelines of AEC Report NUS-531. A base capital-cost model and a base operating- and maintenance-cost model are selected and adjusted for desired size, location, date, etc. Costs are discounted to the year of first commercial operation and levelized to provide annual cost of electric power generation. 3 - Restrictions on the complexity of the problem: The capital cost models are of doubtful validity outside the 500 to 1500 MW(e) range

  4. Normality test for determining the correction factor of isotopic composition in PWR spent fuel

    International Nuclear Information System (INIS)

    Lee, Y. H.; Shin, H. S.; Noh, S. K.; Seo, K. S.

    2001-01-01

    Normality test has been carried out for the ratios of the measured-to-calculated isotopic compositions in PWR spent fuel, using Shapiro-Wilk W, Lilliefors D, Cramer-von Mises and Anderson-Darling. All 38 istopices have been evaluated by means of the 1.5xIQR rule and then outliers have been discarded. As result, it seems that only 20 nuclides are satisfied with the normality at significance level 5 %. 18 Nuclides(samples) including U-235 have higher significance probability(p-value) than 25 % in W-test and p-values obtained by other three tests exceed the upper limit. Besides, in 6 nuclides including Pu-239, it seems that the p-values are between 5 % and 25 % in W test. From these results, in order to predict the isotopic compositions in the conservative point of view, it is decided that the correction factors for the nuclides are determined at the 95/95 probability and confidence level by using tolerance limit-methods with the assumption that only 18 nuclides are satisfied with thr normality

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

    International Nuclear Information System (INIS)

    Ganda, F.; Greenspan, A.

    2005-01-01

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

  6. New technical knowledge to be implemented to the revision of rules on pipe wall thinning management for PWR plants

    International Nuclear Information System (INIS)

    Hirai, Junya; Nakamura, Takao; Amano, Yoichi

    2013-01-01

    Rules for PWR plant pipe wall thinning management were formulated by the Japan Society of Mechanical Engineers in 2006. Since then thinning management of Japanese PWR plants has been carried out based on this rule. Pipe wall thinning phenomena to be dealt with in this rule have been identified in many piping components of power plants. New technical knowledge has been accumulated since the issuance of 2006 edition. We have formulated these knowledge and information about the thinning phenomena in PWR power plants. Given the history of application of this rule, we have to make our best effort to carry out a study of latest technical knowledge and implement them to the revision of rule and improve pipe wall thinning management. This paper summarizes the new technical knowledge and basis to be implemented to the revision of rules on pipe wall thinning management for PWR plants in Japan. (author)

  7. Radial power distribution shaping within a PWR fuel assembly utilizing asymmetrically loaded gadolinia-bearing fuel pins

    International Nuclear Information System (INIS)

    Stone, I.Z.

    1992-01-01

    As in-core fuel management designs evolve to meet the demands of increasing energy output, more innovative methods are developed to maintain power peaking within acceptable thermal margin limits. In-core fuel management staff must utilize various loading pattern strategies such as cross-core movement of fuel assemblies, multibatch enrichment schemes, and burnable absorbers as the primary means of controlling the radial power distribution. The utilization of fresh asymmetrically loaded gadolinia-bearing assemblies as a fuel management tool provides an additional means of controlling the radial power distribution. At Siemens Nuclear Power Corporation (SNP), fresh fuel assemblies fabricated with asymmetrically loaded gadolinia-bearing fuel rods have been used successfully for several cycles of reactor operation. Asymmetric assemblies are neutronically modeled using the same tools and models that SNP uses to model symmetrically loaded gadolinia-bearing fuel assemblies. The CASMO-2E code is used to produce the homogenized macroscopic assembly cross sections for the nodal core simulator. Optimum fuel pin locations within the asymmetrical assembly are determined using the pin-by-pin PDQ7 assembly core model for each new assembly design. The optimum pin location is determined by the rod loading that minimizes the peak-to-average pin power

  8. 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

  9. 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

  10. Parameterized representation of macroscopic cross section in the PWR fuel element considering burn-up cycles

    International Nuclear Information System (INIS)

    Belo, Thiago F.; Fiel, Joao Claudio B.

    2015-01-01

    Nuclear reactor core analysis involves neutronic modeling and the calculations require problem dependent nuclear data generated with few neutron energy groups, as for instance the neutron cross sections. The methods used to obtain these problem-dependent cross sections, in the reactor calculations, generally uses nuclear computer codes that require a large processing time and computational memory, making the process computationally very expensive. Presently, analysis of the macroscopic cross section, as a function of nuclear parameters, has shown a very distinct behavior that cannot be represented by simply using linear interpolation. Indeed, a polynomial representation is more adequate for the data parameterization. To provide the cross sections of rapidly and without the dependence of complex systems calculations, this work developed a set of parameterized cross sections, based on the Tchebychev polynomials, by fitting the cross sections as a function of nuclear parameters, which include fuel temperature, moderator temperature and density, soluble boron concentration, uranium enrichment, and the burn-up. In this study is evaluated the problem-dependent about fission, scattering, total, nu-fission, capture, transport and absorption cross sections for a typical PWR fuel element reactor, considering burn-up cycle. The analysis was carried out with the SCALE 6.1 code package. The results of comparison with direct calculations with the SCALE code system and also the test using project parameters, such as the temperature coefficient of reactivity and fast fission factor, show excellent agreements. The differences between the cross-section parameterization methodology and the direct calculations based on the SCALE code system are less than 0.03 percent. (author)

  11. PWR core and spent fuel pool analysis using scale and nestle

    International Nuclear Information System (INIS)

    Murphy, J. E.; Maldonado, G. I.; St Clair, R.; Orr, D.

    2012-01-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  12. Impact forces on a core shroud of an excited PWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Collard, B.; Vallory, J. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

    2001-07-01

    Seismic excitation of PWR internals may induce large motions of the fuel assemblies (FA). This could result in impact between assemblies or between assemblies and core shroud. Forces generated during these shocks are often the basis for the maximum design loads of the spacer grids and fuel rods. An experimental program has been conducted at the French Nuclear Reactor Directorate (CEA) to measure the impact forces of a reduced scale FA on the test section under different environmental conditions. Within the framework of the tests presented, the effect of the FA environment (air, stagnant water, water under flow) on the maximum impact forces measured at grid levels and on the energy dissipated during the shock is examined. A 'fluid cushioning' effect (dissipative) between the grids and the wall is sought. Experimental results show that the axial flow has a great influence on the impact forces. The greater the axial flow velocity is, the lower the impact forces are. The tests of impact of an assembly on a wall were analyzed compared to the tests carried out without impact. This analysis related on the measured forces of impact and the variation of the measured/computed total energy of the system. The whole of these tests in air and water shows that the 'fluid cushioning' effect required exists but is not significant. Thus the presence of water does not decrease the forces of impact, and does not amplify the quantity of energy dissipated during the shock. The fact that the 'fluid cushioning' effect is weak compared to more analytical tests probably comes from our 'not perfect' or 'realistic' conditions of tests which involve an angle between the grid and the wall at the shock moment.

  13. Parameterized representation of macroscopic cross section in the PWR fuel element considering burn-up cycles

    Energy Technology Data Exchange (ETDEWEB)

    Belo, Thiago F.; Fiel, Joao Claudio B., E-mail: thiagofbelo@hotmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Nuclear reactor core analysis involves neutronic modeling and the calculations require problem dependent nuclear data generated with few neutron energy groups, as for instance the neutron cross sections. The methods used to obtain these problem-dependent cross sections, in the reactor calculations, generally uses nuclear computer codes that require a large processing time and computational memory, making the process computationally very expensive. Presently, analysis of the macroscopic cross section, as a function of nuclear parameters, has shown a very distinct behavior that cannot be represented by simply using linear interpolation. Indeed, a polynomial representation is more adequate for the data parameterization. To provide the cross sections of rapidly and without the dependence of complex systems calculations, this work developed a set of parameterized cross sections, based on the Tchebychev polynomials, by fitting the cross sections as a function of nuclear parameters, which include fuel temperature, moderator temperature and density, soluble boron concentration, uranium enrichment, and the burn-up. In this study is evaluated the problem-dependent about fission, scattering, total, nu-fission, capture, transport and absorption cross sections for a typical PWR fuel element reactor, considering burn-up cycle. The analysis was carried out with the SCALE 6.1 code package. The results of comparison with direct calculations with the SCALE code system and also the test using project parameters, such as the temperature coefficient of reactivity and fast fission factor, show excellent agreements. The differences between the cross-section parameterization methodology and the direct calculations based on the SCALE code system are less than 0.03 percent. (author)

  14. PWR core and spent fuel pool analysis using scale and nestle

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  15. Impact forces on a core shroud of an excited PWR fuel assembly

    International Nuclear Information System (INIS)

    Collard, B.; Vallory, J.

    2001-01-01

    Seismic excitation of PWR internals may induce large motions of the fuel assemblies (FA). This could result in impact between assemblies or between assemblies and core shroud. Forces generated during these shocks are often the basis for the maximum design loads of the spacer grids and fuel rods. An experimental program has been conducted at the French Nuclear Reactor Directorate (CEA) to measure the impact forces of a reduced scale FA on the test section under different environmental conditions. Within the framework of the tests presented, the effect of the FA environment (air, stagnant water, water under flow) on the maximum impact forces measured at grid levels and on the energy dissipated during the shock is examined. A 'fluid cushioning' effect (dissipative) between the grids and the wall is sought. Experimental results show that the axial flow has a great influence on the impact forces. The greater the axial flow velocity is, the lower the impact forces are. The tests of impact of an assembly on a wall were analyzed compared to the tests carried out without impact. This analysis related on the measured forces of impact and the variation of the measured/computed total energy of the system. The whole of these tests in air and water shows that the 'fluid cushioning' effect required exists but is not significant. Thus the presence of water does not decrease the forces of impact, and does not amplify the quantity of energy dissipated during the shock. The fact that the 'fluid cushioning' effect is weak compared to more analytical tests probably comes from our 'not perfect' or 'realistic' conditions of tests which involve an angle between the grid and the wall at the shock moment

  16. A comparison study on radioactive waste management effectiveness in various nuclear fuel cycles

    International Nuclear Information System (INIS)

    Ko, Won Il; Kim, Ho Dong

    2001-07-01

    This study examines whether the DUPIC (Direct Use of Spent PWR Fuel In CANDU) fuel cycle make radioactive waste management more effective, by comparing it with other fuel cycles such as the PWR (Pressurized Water Reactor) once-through cycle, the HWR (Pressurized Heavy Water Reactor) once-through cycle and the thermal recycling option to use an existing PWR with MOX (Mixed Oxide) fuel. This study first focuses on the radioactive waste volume generated in all fuel cycle steps, which could be one of the measures of effectiveness of the waste management. Then the total radioactive waste disposition cost is estimated based on two units measuring; m3/GWe-yr and US$/GWe-yr. We find from the radioactive waste volume estimation that the DUPIC fuel cycle could have lower volumes for milling tailings, low level waste and spent fuel than those of other fuel cycle options. From the results of the disposition cost analysis, we find that the DUPIC waste disposition cost is the lowest among fuel cycle options. If the total waste disposition cost is used as a proxy for quantifying the easiness or difficulty in managing wastes, then the DUPIC option actually make waste management easier

  17. Quantitative Analysis of Kr-85 Fission Gas Release from Dry Process for the Treatment of Spent PWR Fuel

    International Nuclear Information System (INIS)

    Park, Geun Il; Cho, Kwang Hun; Lee, Dou Youn; Lee, Jung Won; Park, Jang Jin; Song, Kee Chan

    2007-01-01

    As spent UO 2 fuel oxidizes to U 3 O 8 by air oxidation, a corresponding volume expansion separate grains, releasing the grain-boundary inventory of fission gases. Fission products in spent UO 2 fuel can be distributed in three major regions : the inventory in fuel-sheath gap, the inventory on grain boundaries and the inventory in UO 2 matrix. Release characteristic of fission gases depends on its distribution amount in three regions as well as spent fuel burn-up. Oxidation experiments of spent fuel at 500 .deg. C gives the information of fission gases inventory in spent fuel, and further annealing experiments at higher temperature produces matrix inventory of fission gases on segregated grain. In previous study, fractional release characteristics of Kr- 85 during OREOX (Oxidation and REduction of Oxide fuel) treatment as principal key process for recycling spent PWR fuel via DUPIC cycle have already evaluated as a function of fuel burn-up with 27.3, 35 and 65 MWd/tU. In this paper, new release experiment results of Kr-85 using spent fuel with burn- up of 58 GWd/tU are included to evaluate the fission gas release behavior. As a point of summary in fission gases release behavior, the quantitative analysis of Kr- 85 release characteristics from various spent fuels with different burn-up during voloxidation and OREOX process were reviewed

  18. Supercritical water-cooled reactor fuel management and economic comparison and analysis

    International Nuclear Information System (INIS)

    Cai Guangming; Ruan Liangcheng; Liu Xuechun

    2014-01-01

    The supercritical water-cooled reactor (SCWR) is expected to have an excellent fuel economical efficiency because of its high thermal efficiency. This article compares CSR1OOO with the current mainstream PWR and ABWR on the aspect of the economical efficiency of fuel management, and finally makes an unexpected conclusion that the SCWR has worse fuel economy than others. And it remains to be deliberated whether the SCWR will be the fourth generation of nuclear system. (authors)

  19. 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.)

  20. 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

  1. 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)

  2. 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)

  3. A proposal for a unified fuel thermal conductivity model available for UO{sub 2}, (U-Pu)O{sub 2} and UO{sub 2}-GD{sub 2}O{sub 3} PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Baron, D [Electrice de France, Moret-sur-Loing (France); Couty, J C [Electricite de France (EDF), 69 - Villeurbanne (France)

    1997-08-01

    In order to cope with the current fuel management targets which are focussed on higher discharge burnups, initial {sup 235}U fuel enrichments have been increased from 3.25% to 4%. To avoid an increase in boron concentration in the primary circuit, Gadolinium is used as a burnable poison, spread in the uranium oxide matrix of selected rods, in order to absorb the initial reactivity excess. Obviously, fuel thermal conductivity is affected when introducing any stranger element. Previously, the EDF thermomechanical code provided two different models to simulate the fuel thermal conductivity: one available for UO{sub 2} and (U-Pu)O{sub 2} fuels, the other for Gadolinia fuels, depending on the calculations to be done. No effect of the initial fuel stoichiometry was taken into account in the second model. That situation suggested the development of a unified model available for any fuels presently loaded in the EDF PWR reactors. This paper deals with the choice of the formulation, the data base used and the methodology applied for parameter fitting. Results in terms of measured versus predicted evaluation are then discussed. (author). 11 refs, 5 figs.

  4. Aging management of PWR reactor internals in U.S. plants

    International Nuclear Information System (INIS)

    Amberge, K.J.; Demma, A.

    2015-01-01

    This paper describes the development, aging management strategies and inspection results of the Pressurized Water Reactor (PWR) vessel internals inspection and evaluation guidelines. The goal of these guidelines is to provide PWR owners with robust aging management strategies to monitor degradation of internals components to support life extension as well as the current period of operation and power up-rate activities. The implementation of these guidelines began in 2010 within the U.S. PWR fleet and several examinations have been performed since. Examples of inspection results are presented for selected vessel internals components and are compared with simulation results. In summary, to date there have been no observations of austenitic stainless steel stress corrosion cracking (SCC), which is consistent with expectations based on the current understanding of the mechanism. Observations of irradiation assisted stress corrosion cracking (IASCC) have been limited and only found in baffle former bolting. Additionally, no macroscopic effects or global observations of void swelling impacts on general conditions of reactor internal hardware have been observed. (authors)

  5. Dynamic strain aging of zircaloy-4 PWR fuel cladding in biaxial stress state

    International Nuclear Information System (INIS)

    Park, Ki Seong; Lee, Byong Whi

    1989-01-01

    The expanding copper mandrel test performed at three strain rates (3.2x10E-5/s,2.0x10E-6/s and 1.2x10E-7/s) over 553-873 K temperature range by varying the heating rates (8-10deg C/s,1-2deg C/s and 0.5deg C/s) in air and in vacuum (5x10E-5 torr). The yield stress peak, the strain rate sensitivity minimum and the activation volume peaks could be explained in terms of the dynamic strain aging. The activation energy for dynamic strain aging obtained from the yield stress peak temperature and strain rate was 196 KJ/mol and this value was in good agreement with the activation energy for oxygen diffusion in α-zirconium and Zircaloy-2 (207-220KJ/mol). Therefore, oxygen atoms are responsible for the dynamic strain aging which appeared between 573K and 673K. The yield stress increase due to the oxidation was obtained by comparing the yield stress in air with that in vacuum and represented by the percentage increase of yield stress (σ y a -σ y v /σ y v ). The slower the strain rate, the greater the percentage increase occurs. In order to estimate the yield stress of PWR fuel cladding material under the service environment, the yield stress in water was obtained by comparing the oxidation rate in air that in water assuming the relationship between the oxygen pick-up amount and the yield stress increase. (Author)

  6. TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

    International Nuclear Information System (INIS)

    DOE

    1997-01-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k eff , of a spent nuclear fuel package. Fifty-seven UO 2 , UO 2 /Gd 2 O 3 , and UO 2 /PuO 2 critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k eff (which can be a function of the trending parameters) such that the biased k eff , when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection

  7. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package

  8. 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)

  9. 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)

  10. Nuclear fuel management optimization for LWRs

    International Nuclear Information System (INIS)

    Turinsky, Paul J.

    1997-01-01

    LWR in core nuclear fuel management involves the placement of fuel and control materials so that a specified objective is achieved within constraints. Specifically, one is interested in determining the core loading pattern (LP of fuel assemblies and burnable poisons and for BWR, also control rod insertion versus cycle exposure. Possible objectives include minimization of feed enrichment and maximization of cycle energy production, discharge burnup or thermal margin. Constraints imposed relate to physical constraints, e.g. no discrete burnable poisons in control rod locations, and operational and safety constraints, e.g. maximum power peaking limit. The LP optimization problem is a large scale, nonlinear, mixed-integer decision variables problem with active constraints. Even with quarter core symmetry imposed, there are above 10 100 possible LPs. The implication is that deterministic optimization methods are not suitable, so in this work we have pursued using the stochastic Simulated Annealing optimization method. Adaptive penalty functions are used to impose certain constraints, allowing unfeasible regions of the search space to be transverse. Since ten of thousands of LPs must be examined to achieve high computational efficiency, higher-order Generalized Perturbation Theory is utilized to solve the Nodal Expansion Method for of the two-group neutron diffusion. These methods have been incorporated into the FORMOSA series of codes and used to optimize PWR and BWR reload cores. (author). 9 refs., 3 tabs

  11. 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

  12. 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.

  13. 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

  14. Use of 'tail' as spent fuel dilution factor of Angra-1 (PWR) for use in the Embalse (Candu) reactor

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Maiorino, Jose Rubens

    1995-01-01

    This work purposes a process to use the tail of isotopic enrichment as a factor of dilution (blending) for the burned fuel of Angra-I reactor (PWR) for final utilization in the Embalse (Candu). It was made use of the same technic in previous works that used natural uranium. For this purpose, it was made a tail parametrization inside of the traditional limits of enrichment (between 0.2 and 0.3%). The study showed that the tail utilization represents great savings for the uranium supplies and environment and economic advantages. (author). 8 refs, 4 figs, 11 tabs

  15. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL) [de

  16. An expert system for PWR core operation management

    Energy Technology Data Exchange (ETDEWEB)

    Ida, Toshio; Masuda, Masahiro; Nishioka, Hiromasa

    1988-01-01

    Planning for restartup after planned or unplanned reactor shutdown and load-follow operations is an important task in the core operation management of pressurized water reactors (PWRs). These planning problems have been solved by planning experts using their expertise and the computational prediction of core behavior. Therefore, the quality of the plan and the time consumed in the planning depend heavily on the skillfulness of the planning experts. A knowledge engineering approach has been recently considered as a promising means to solve such complicated planning problems. Many knowledge-based systems have been developed so far, and some of them have already been applied because of their effectiveness. The expert system REPLEX has been developed to aid core management engineers in making a successful plan for the restartup or the load-follow operation of PWRs within a shorter time. It can maintain planning tasks at a high-quality level independent of the skillfulness of core management engineers and enhance the efficiency of management. REPLEX has an explanation function that helps user understanding of plans. It could be a useful took, therefore, for the training of core management engineers.

  17. An expert system for PWR core operation management

    International Nuclear Information System (INIS)

    Ida, Toshio; Masuda, Masahiro; Nishioka, Hiromasa.

    1988-01-01

    Planning for restartup after planned or unplanned reactor shutdown and load-follow operations is an important task in the core operation management of pressurized water reactors (PWRs). These planning problems have been solved by planning experts using their expertise and the computational prediction of core behavior. Therefore, the quality of the plan and the time consumed in the planning depend heavily on the skillfulness of the planning experts. A knowledge engineering approach has been recently considered as a promising means to solve such complicated planning problems. Many knowledge-based systems have been developed so far, and some of them have already been applied because of their effectiveness. The expert system REPLEX has been developed to aid core management engineers in making a successful plan for the restartup or the load-follow operation of PWRs within a shorter time. It can maintain planning tasks at a high-quality level independent of the skillfulness of core management engineers and enhance the efficiency of management. REPLEX has an explanation function that helps user understanding of plans. It could be a useful took, therefore, for the training of core management engineers

  18. Ageing management database development for PWR NPP steam generator

    International Nuclear Information System (INIS)

    Liu Hongyun; Xu Liangjun; Xiong Changhuai; Wang Xianyuan

    2005-01-01

    Steam generator (SG) is one of the key safe important equipment of NPP, which is covered by NPP aging management program. Steam Generator Aging Management Dabatase (SGAMDB) is developed to provide necessary information for SG aging management. RINPO is developing SGAMDB for domestic NPP. This system contains information and data about SG design, manufacture, operation and maintenance. The information include NPP fundamental data, SG design data, SG aging mechanism, SG operation data, SG ISI data, SG maintenance data and SG evaluation interface. The system runs at the intranet of Qinshan-1 NPP with B/S mode. It can provide information inquire and fundamental analysis for NPP SG aging team and SG aging researcher's. In addition, it provides necessary information and data for SG aging analysis and evaluation, such as all pressure test process and flaws of tubes, and collects the analysis results. (authors)

  19. 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.

  20. Safety Aspects of Radioactive Waste Management in Different Nuclear Fuel Cycle Policies, a Comparative Study

    International Nuclear Information System (INIS)

    Gad Allah, A.A.

    2009-01-01

    With the increasing demand of energy worldwide, and due to the depletion of conventional natural energy resources, energy policies in many countries have been devoted to nuclear energy option. On the other hand, adopting a safe and reliable nuclear fuel cycle concept guarantees future nuclear energy sustain ability is a vital request from environmental and economic point of views. The safety aspects of radioactive waste management in the nuclear fuel cycle is a topic of great importance relevant to public acceptance of nuclear energy and the development of nuclear technology. As a part of nuclear fuel cycle safety evaluation studies in the department of nuclear fuel cycle safety, National Center for Nuclear Safety and Radiation Control (NCNSRC), this study evaluates the radioactive waste management policies and radiological safety aspects of three different nuclear fuel cycle policies. The once-through fuel cycle (OT- fuel cycle) or the direct spent fuel disposal concept for both pressurized light water reactor ( PWR) and pressurized heavy water reactor (PHWR or CANDU) systems and the s elf-generated o r recycling fuel cycle concept in PWR have been considered in the assessment. The environmental radiological safety aspects of different nuclear fuel cycle options have been evaluated and discussed throughout the estimation of radioactive waste generated from spent fuel from these fuel cycle options. The decay heat stored in the spent fuel was estimated and a comparative safety study between the three fuel cycle policies has been implemented

  1. 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

  2. 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)

  3. 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

  4. 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.)

  5. Assessment of management alternatives for LWR wastes. Volume 3. Description of German scenarios for PWR and BWR wastes

    International Nuclear Information System (INIS)

    Santraille, S.

    1993-01-01

    This report deals with the description of a management route for PWR waste relying to a certain extent on German practices in this particular area. This description, which aims at providing input data for subsequent cost evaluation, includes all management steps which are usually implemented for solid, liquid and gaseous wastes from their production up to the interim storage of the final waste products. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for PWR and BWR wastes based on economical and radiological criteria

  6. Management routes for materials arising from the decommissioning of a PWR reactor

    International Nuclear Information System (INIS)

    Klein, M.; Demeulemeester, Y.; Moers, S.; Ponnet, M.

    2001-01-01

    The management of wastes from decommissioning is described for the on-going dismantling of the BR3 PWR small reactor. The incentive is put on the radionuclides characterization, the description of the various waste streams, the conditioning techniques for low radioactive waste (LAW) to high radioactive waste (RAW), the alternative evacuation routes (recycling in the nuclear, free release by decontamination) and the minimization of secondary wastes during dismantling. Finally, some considerations are given on the overall dismantling cost and on the relative costs of the various evacuation routes. (author)

  7. Selection of detailed items for periodic safety review on PWR radwaste management system

    Energy Technology Data Exchange (ETDEWEB)

    Sung, K. B.; Ahn, Y. S.; Park, Y. S.; Kim, S. H.; Kim, J. T. [Korea Hydric and Nuclear Power Company, Taejon (Korea, Republic of)

    2003-10-01

    Selection of detailed-items for Periodic Safety Review on PWR radwaste management system, the main component could be faithfully clarified according to the purpose of establishment on each system and basic purpose. It is proper to select detailed-items those of radioactivities in the reactor coolant activity levels and the released volume of liquid and gaseous radioactive material on safety performance. It's also proper to select solid radwaste production quantities as detailed-item that it would be predict the next ten years trends after PSR.

  8. 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)

  9. Modelling of pellet cladding interaction during power ramps in PWR rods by means of Transuranus fuel rod analysis code

    International Nuclear Information System (INIS)

    Di Marcello, V.; Luzzi, L.

    2008-01-01

    Pellet-cladding interaction (PCI) in PWR type rods subjected to power ramps was analysed by means of TRANSURANUS (TU) fuel rod performance code. PCI phenomena depend on the fuel power history - i.e. by several irradiation and thermal induced phenomena occurring in the fuel rod and mutually interacting during its life in reactor - and may become critical for cladding integrity under accidental conditions. Ten test fuel rods, whose power histories and post irradiation experiment (PIE) data were available from the OECD/NEA-IAEA International Fuel Performance Experiment (UTE) database through the Studsvik SUPER-RAMP Project, were simulated by TRANSURANUS. During a power ramp pellet gaseous swelling can be inhibited by cladding pressure and can be over-predicted by a normal operation swelling model. This phenomenon was simulated by a new formulation of a fuel swelling model already available in the code, in order to consider hot pressing of inter-granular -fuel porosity due to the high hydrostatic stress resulting from PCI: it was found that TRANSURANUS, as a result of the proposed swelling formulation as well as of the accurate modelling of the other phenomena occurring during irradiation, gives correct predictions on PCI induced fuel rod failures. In addition, PCI failure threshold identified by TRANSURANUS was compared with the technological limits known in literature: the possibility of relaxing these limits for low burn-up values and the preponderance of the European fuel rod design in front of PCI emerged from TU analyses. Finally, a good agreement was found between TU evaluations and PIE data, with regard to fission gas release, fuel grain growth, and creep, corrosion and elongation of the cladding. (authors)

  10. Uranium resource utilization improvements in the once-through PWR fuel cycle

    International Nuclear Information System (INIS)

    Matzie, R.A.

    1980-04-01

    In support of the Nonproliferation Alternative Systems Assessment Program (NASAP), Combustion Engineering, Inc. performed a comprehensive analytical study of potential uranium utilization improvement options that can be backfit into existing PWRs operating on the once-through uranium fuel cycle. A large number of potential improvement options were examined as part of a preliminary survey of candidate options. The most attractive of these, from the standpoint of uranium utilization improvement, economic viability, and ease of implementation, were then selected for detailed analysis and were included in a single composite improvement case. This composite case represents an estimate of the total savings in U 3 O 8 consumption that can be achieved in current-design PWRs by implementing improvements which can be developed and demonstrated in the near term. The improvement options which were evaluated in detail and included in the composite case were a new five-batch, extended-burnup fuel management scheme, low-leakage fuel management, modified lattice designs, axial blankets, reinsertion of initial core batches, and end-of-cycle stretchout

  11. 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.

  12. Fuel management of HTR-10

    International Nuclear Information System (INIS)

    Wu Zongxin; Jing Xingqing

    2001-01-01

    The 10 MW high temperature cooled reactor (HTR-10) built in Tsinghua University is a pebble bed type of HTGR. The continuous recharge and multiple-pass of spherical fuel elements are used for fuel management. The initiative stage of core is composed of the mix of spherical fuel elements and graphite elements. The equilibrium stage of core is composed of identical spherical fuel elements. The fuel management during the transition from the initiative stage to the equilibrium stage is a key issue for HTR-10 physical design. A fuel management strategy is proposed based on self-adjustment of core reactivity. The neutron physical code is used to simulate the process of fuel management. The results show that the graphite elements, the recharging fuel elements below the burn-up allowance, and the discharging fuel elements over the burn-up allowance could be identified by burn-up measurement. The maximum of burn-up fuel elements could be controlled below the burn-up limit

  13. Source Term Characteristics Analysis for Structural Components in PWR spent fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kook, Dong Hak; Choi, Heui Joo; Cho, Dong Keun [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core under different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be 1.32x1015 Bequerels, 238 Watts, 4.32x109 m3 water, respectively, at 10 years after discharge. Those values correspond to 0.6 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 25{approx}50 % and 35{approx}40 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important

  14. Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE

    Energy Technology Data Exchange (ETDEWEB)

    Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.; Di-Salvo, J.; Antony, M.; Pepino, A. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

    2012-07-01

    The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now. The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)

  15. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J L; Lopez, J

    1986-07-01

    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.

  16. Study of transient heat transfer in a fuel rod 3D, in a situation of unplanned shutdown of a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Martins, Rodolfo Ienny; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: rodolfoienny@gmail.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The study, in situations involving accidents, of heat transfer in fuel rods is of known importance, since it can be used to predict the temperature limits in designing a nuclear reactor, to assist in making more efficient fuel rods, and to increase the knowledge about the behavior of the reactor's components, a crucial aspect for safety analysis. This study was conducted using as parameter the fuel rod that has the highest average power in a typical PWR reactor. For this, we developed a program (Fuel{sub R}od{sub 3}D) in Fortran language using the Finite Elements Method (FEM) for the discretization of a fuel rod and coolant channel, in order to study the temperature distribution in both the fuel rod and the coolant channel. Transient parameters were coupled to the heat transfer equations in order to obtain details of the behavior of the rod and the channel, which allows the analysis of the temperature distribution and its change over time. This work aims to present a study case of an accident where there is a lack of energy in the reactor's coolant pumps and in the diesel engines, resulting in an unplanned shutdown of the reactor. In order to achieve the intended goal, the present work was divided as follows: a short introduction about heat transfer, including the equations concerning the fuel rod and the energy equation in the channel, an explanation about how the verification of the Fuel{sub R}od{sub 3}D program was made, and the analysis of the results. (author)

  17. High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

    Science.gov (United States)

    Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

    2014-08-01

    In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover

  18. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    International Nuclear Information System (INIS)

    1996-01-01

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B 4 C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B 4 C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B 4 C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H 2 generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B 4 C were not treated. In general the confidence in code predictions decreases with processing core damage. (N.T.)

  19. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B{sub 4}C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B{sub 4}C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B{sub 4}C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H{sub 2} generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B{sub 4}C were not treated. In general the confidence in code predictions decreases with processing core damage. 36 refs.

  20. Introduction of HTR-PM Operation and Fuel Management System

    International Nuclear Information System (INIS)

    Liu Fucheng; Luo Yong; Gao Qiang

    2014-01-01

    There is a big difference between High Temperature Gas-cooled Reactor Pebble-modules Demonstration Project(HTR-PM) and PWR in operation mode. HTR-PM is a continually refuelled reactor, and the operation and fuel management of it, which affect each other, are inseparable. Therefore, the analysis of HTR-PM fuel management needs to be carried out “in real time”. HTR-PM operation and fuel management system is developed for on-power refuelling mode of HTR-PM. The system, which calculates the core neutron flux and power distribution, taking high-temperature reactor physics analysis software-VSOP as a basic tool, can track and predict the core state online, and it has the ability to restructure core power distribution online, making use of ex-core detectors to correct and check tracking calculation. Based on the ability to track and predict, it can compute the core parameters to provide support for the operation of the reactor. It can also predict the operation parameters of the reactor to provide reference information for the fuel management.The contents of this paper include the development purposes, architecture, the main function modules, running process, and the idea of how to use the system to carry out HTR-PM fuel management. (author)

  1. Quality management of nuclear fuel

    International Nuclear Information System (INIS)

    2006-01-01

    The Guide presents the quality management requirements to be complied with in the procurement, design, manufacture, transport, receipt, storage, handling and operation of nuclear fuel. The Guide also applies to control rods and shield elements to be placed in the reactor. The Guide is mainly aimed for the licensee responsible for the procurement and operation of fuel, for the fuel designer and manufacturer and for other organisations, whose activities affect fuel quality and the safety of fuel transport, storage and operation. General requirements for nuclear fuel are presented in Section 114 of the Finnish Nuclear Energy Decree and in Section 15 of the Government Decision (395/1991). Regulatory control of the safety of fuel is described in Guides YVL6.1, YVL6.2 and YVL6.3. An overview of the regulatory control of nuclear power plants carried out by STUK (Radiation and Nuclear Safety Authority, Finland) is clarified in Guide YVL1.1

  2. 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)

  3. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant

    International Nuclear Information System (INIS)

    Monteiro, Iara Arraes

    1999-02-01

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  4. Refitting of the 'Celimene' hot cell for following up the fuel assembly of 900 MWe PWR power reactors

    International Nuclear Information System (INIS)

    Lhermenier, Andre; Van Craeynest, J.-C.

    1980-05-01

    The 'Celimene' cell adjoining the EL3 reactor provides for the acceptance, handling and the examination of irradiated fuel assemblies from power reactors (length approximately 4m, weight approximately 700 kg). Within the framework of the PWR fuel behavior follow-up or reprocessing, it is possible to extract an assembly representative of the normal fuel cycle, carry out non destructive tests on this assembly, extract pencils from it and re-insert this assembly, after refitting the head, into the normal fuel cycle for handling in a reprocessing plant or storage pond. Given suitable refitting techniques, the re-irradiation of the assembly can be considered after examination. Significant changes have been made to the buildings and the hoist facilities for handling very heavy flasks. It was necessary to rearrange the handling, machining and in-cell storage facilities. The development of an inspection rig will make it possible, some time in 1980, to carry out non destructive tests of assemblies, optical and metrological examination of assemblies prior to dismantling or of the structure after dismantling [fr

  5. 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

  6. Rod consolidation of RG and E's [Rochester Gas and Electric Corporation] spent PWR [pressurized water reactor] fuel

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1987-05-01

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister

  7. Management of spent fuel in Republic of Korea

    International Nuclear Information System (INIS)

    Pak, Hyun-Soo; Seo, In-Seok; Pak, Sang-Ki.

    1989-01-01

    At present in Republic of Korea, 8 PWR and 1 CANDU power plants are in operation or under construction, and the total capacity of power generation facilities has become 7.6 GWe. In addition, two PWRs of more than 900 MWe each are expected to be constructed by mid 1990s. More than 50 % of the electric power demand was supplied by nuclear power generation since 1987, but the spent fuel generated in nuclear power plants is stored in storage water tanks in respective reactor sites. The total capacity of spent fuel to be stored in the AR facilities of 9 nuclear power plants is about 2730 MTU, and the spent fuel released from these reactors since 1980 is about 810 MTU. The present capacity of AR storage pools seems to be used up by mid 1990s. According to the revised Atomic Energy Acts in May, 1986, the government is to take the responsibility of spent fuel management, and the policy of constructing the storage facilities outside reactor sites by the end of 1997 was established by the Atomic Energy Commission. The responsibility of the management of spent fuel that exceeds the present capacity of AR pools is to be taken by KEPCO, therefore the preliminary analysis of the feasible option on the extension of AR facilities and the comprehensive management plan for spent fuel placing emphasis on the research and development of away-from-reactor storage were decided. (Kako, I.)

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

    International Nuclear Information System (INIS)

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

    1981-04-01

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

  9. AGR fuel management using PANTHER

    International Nuclear Information System (INIS)

    Haddock, S.A.; Parks, G.T.

    1995-01-01

    This paper describes recent improvements in the AGR fuel management methodology implemented within PANTHER and the use of the code both for stand-alone calculations and within an automatic optimisation procedure. (author)

  10. Spent fuel management in Spain

    International Nuclear Information System (INIS)

    Gonzalez, J.L.

    2002-01-01

    The spent fuel management strategy in Spain is presented. The strategy includes temporary solutions and plans for final disposal. The need for R and D including partitioning and transmutation, as well as the financial constraints are also addressed. (author)

  11. In-core fuel management programs for nuclear power reactors

    International Nuclear Information System (INIS)

    1984-10-01

    In response to the interest shown by Member States, the IAEA organized a co-ordinated research programme to develop and make available in the open domain a set of programs to perform in-core fuel management calculations. This report summarizes the work performed in the context of the CRP. As a result of this programme, complete in-core fuel management packages for three types of reactors, namely PWR's, BWR's and PHWR are now available from the NEA Data Bank. For some reactor types, these program packages are available with three levels of sophistication ranging from simple methods for educational purposes to more comprehensive methods that can be used for reactor design and operation. In addition some operating data have been compiled to allow code validation. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, T.; Iwasaki, T.; Wada, K. [Tohoku Univ., Graduate School of Engineering, Dept. of Quantum Science and Energy Engineering, Sendai 980-8579 (Japan); Suyama, K. [Japan Atomic Energy Agency, Shirakata-Shirane 2-4, Naka-gun, Ibaraki-ken 319-1195 (Japan)

    2006-07-01

    To examine the procedures of the reprocessing, the vitrification and the geologic disposal, precise burn-up calculation for high burn-up and MOX fuels has been performed for not only PWR but also BWR by using SWAT and SWAT2 codes which are the integrated bum-up calculation code systems combined with the bum-up calculation code, ORIGEN2, and the transport calculation code, SRAC (the collision probability method) or MVP (the continuous energy Monte Carlo method), respectively. The calculation results shows that all of the evaluated items (heat generation and concentrations of Mo and Pt) largely increase and those significantly effect to the current procedures of the vitrification and the geologic disposal. The calculation result by SWAT2 confirms that the bundle calculation is required for BWR to be discussed about those effects in details, especially for the MOX fuel. (authors)

  13. Development of base technology for high burnup PWR fuel improvement Volume 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Eun; Lee, Sang Hee; Bae, Seong Man [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Chung, Jin Gon; Chung, Sun Kyo; Kim, Sun Du [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Kim, Jae Won; Chung, Sun Kyo; Kim, Sun Du [Korea Nuclear Fuel Development Inst., Seoul (Korea, Republic of)

    1995-12-31

    Development of base technology for high burnup nuclear fuel -Development of UO{sub 2} pellet manufacturing technology -Improvement of fuel rod performance code -Improvement of plenum spring design -Study on the mechanical characteristics of fuel cladding -Organization of fuel failure mechanism Establishment of next stage R and D program (author). 226 refs., 100 figs.

  14. Fuel cycle management in Finland

    International Nuclear Information System (INIS)

    Vaeyrynen, H.; Mikkola, I.

    1987-01-01

    Both Finnish utilities producing nuclear power - Imatran Voima Oy (IVO) and Teollisuuden Voima Oy (Industrial Power Co. Ltd, TVO) - have created efficient fuel cycle management systems. The systems however differ in almost all respects. The reason is that the principal supplier for IVO is the Soviet Union and for TVO is Sweden. A common feature of both systems at the front end of the cycle is the building of stockpiles in order to provide for interruptions in fuel deliveries. Quality assurance supervision at the fuel factory for IVO is regulated by the Soviet Chamber of Commerce and Industry and a final control is made in Finland. The in-core fuel management is done by IVO using codes developed in Finland. The whole IVO fuel cycle is basically a leasing arrangement. The spent fuel is returned to the USSR after five years cooling. TVO carries out the in-core fuel management using a computer code system supplied by Asea-Atom. TVO is responsable for the back end of the cycle and makes preparations for the final disposal of the spent fuel in Finland. 6 refs., 2 figs

  15. Spent fuel management in Japan

    International Nuclear Information System (INIS)

    Mineo, H.; Nomura, Y.; Sakamoto, K.

    1998-01-01

    In Japan 52 commercial nuclear power units are now operated, and the total power generation capacity is about 45 GWe. The cumulative amount of spent fuel arising is about 13,500 tU as of March 1997. Spent fuel is reprocessed, and recovered nuclear materials are to be recycled in LWRs and FBRs. In February 1997 short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, backend measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away from reactor sites, considering the increasing amount of spent fuel arising. Research and development on spent fuel storage has been carried out, particularly on dry storage technology. Fundamental studies are also conducted to implement the burnup credit into the criticality safety design of storage and transportation casks. Rokkasho reprocessing plant is being constructed towards its commencement in 2003, and Pu utilization in LWRs will be started in 1999. Research and development of future recycling technology are also continued for the establishment of nuclear fuel cycle based on FBRs and LWRs. (author)

  16. Fuel reprocessing and waste management

    International Nuclear Information System (INIS)

    Philippone, R.L.; Kaiser, R.A.

    1989-01-01

    Because of different economic, social and political factors, there has been a tendency to compartmentalize the commercial nuclear power industry into separate power and fuel cycle operations to a greater degree in some countries compared to other countries. The purpose of this paper is to describe how actions in one part of the industry can affect the other parts and recommend an overall systems engineering approach which incorporates more cooperation and coordination between individual parts of the fuel cycle. Descriptions are given of the fuel cycle segments and examples are presented of how a systems engineering approach has benefitted the fuel cycle. Descriptions of fuel reprocessing methods and the waste forms generated are given. Illustrations are presented describing how reprocessing options affect waste management operations and how waste management decisions affect reprocessing

  17. The irradiation performance of austenitic stainless steel clade PWR fuel rods

    International Nuclear Information System (INIS)

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

    1988-01-01

    The steady state irradiation performance of austenitic stainless steel clad pressurized water reactor fuel rods is modeled with fuel performance codes of the FRAP series. These codes, originally developed to model the thermal-mechanical behavior of zircaloy clad fuel rods, are modified to model stainless steel clad fuel rods. The irradiation thermal-mechanical behavior of type 348 stainless steel and zircaloy fuel rods is compared. (author) [pt

  18. Long-Term Dry Storage of High Burn-Up Spent Pressurized Water Reactor (PWR) Fuel in TAD (Transportation, Aging, and Disposal) Containers

    International Nuclear Information System (INIS)

    Hwang, Yong Soo

    2008-12-01

    A TAD canister, in conjunction with specially-designed over-packs can accomplish the functions of transportation, aging, and disposal (TAD) in the management of spent nuclear fuel (SNF). Industrial dry cask systems currently available for SNF are licensed for storage-only or for dual-purpose (i.e., storage and transportation). By extending the function to include the indefinite storage and perhaps, eventual geologic disposal, the TAD canister would have to be designed to enhance, among others, corrosion resistance, thermal stability, and criticality-safety control. This investigative paper introduces the use of these advanced iron-based, corrosion-resistant materials for SNF transportation, aging, and disposal.The objective of this investigative project is to explore the interest that KAERI would research and develop its specific SAM coating materials for the TAD canisters to satisfy the requirements of corrosion-resistance, thermal stability, and criticality-controls for long-term dry storage of high burn-up spent PWR fuel

  19. Results and comments on the gamma spectrometry examinations carried out on PWR and fast neutron fuel elements

    International Nuclear Information System (INIS)

    Pineira, Thomas; Mouchnino, Michel; Juste, Guy; Vignesoult, Nicole.

    1980-05-01

    The gamma spectrometry analyses on the fuel elements of PWR and fast neutron systems have experienced a significant growth in the CEA. This nondestructive, quick, inexpensive and quantitative method, seems to us particularly advantageous for qualifying the behavior of fuel under irradiation. However, in order to use it to the maximum, it must have reached a high degree of automation and the interpretation of the results must be the outcome of a coherent team that includes gamma spectrometry and fuel element specialists, since the growth of hot cell gamma spectrometry involves the processing of a considerable number of data upon which the quality of the results depends (large number of spectra per pencil analyzed, dimension of the 2000 or 4000 channel spectra, number of lines studied, etc.). Therefore the need to make the most of the information and, in particular, to present the results in a form suitable for direct processing in a minimum response time, requires a highly automated system. Further, the more specific results of gamma spectrometry correlated to the metallurgical results obtained in the laboratories should contribute efficiently to obtaining major information [fr

  20. Duo_2-Steel cermet manufacturing technology for PWR Spent Nuclear Fuel (SNF) casks

    International Nuclear Information System (INIS)

    Siti Alimah; Budiarto

    2005-01-01

    Assessment of DUO_2-Steel cermet manufacturing technology for PWR SNF casks has been done. DUO_2-Steel cermet consisting of DUO_2 particulates and other particulates, embedded in a steel matrix. Cermet SNF casks have the potential for superior performance compared with casks constructed of other materials. The addition of DUO_2 ceramic particulates can increase SNF cask capacity, improve of repository performance and disposal of excess depleted uranium as potential waste. Two sets of cermet manufacturing technologies are casting and powder metallurgy. Three casting methods are infusion casting, traditional casting and centrifugal casting. While for powder metallurgy methods there are traditional method and new method. DUO_2-Steel cermet have traditionally been produced by powder metallurgy methods. The production of a cask, however, presents special requirements: the manufacture of an annular object with weights up to 100 tons, and methods are being not to manufacture a cermet of this size and geometry. A new powder metallurgy method, is a method for manufacturing cermet for PWR SNF cask. This powder metallurgy techniques have potentials low costs and provides greater freedom In the design of the cermet cask by allowing variable cermet properties. (author)

  1. 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)

  2. Calculation of burnup and power dependence on fission gas released from PWR type reactor fuel element

    International Nuclear Information System (INIS)

    Edy-Sulistyono

    1996-01-01

    Burn up dependence of fission gas released and variation power analysis have been conducted using FEMXI-IV computer code program for Pressure Water Reactor Fuel During steady-state condition. The analysis result shows that the fission gas release is sensitive to the fuel temperature, the increasing of burn up and power in the fuel element under irradiation experiment

  3. Spent fuel management overview: a global perspective

    International Nuclear Information System (INIS)

    Bonne, A.; Crijns, M.J.; Dyck, P.H.; Fukuda, K.; Mourogov, V.M.

    1999-01-01

    The paper defines the main spent fuel management strategies and options, highlights the challenges for spent fuel storage and gives an overview of the regional balances of spent fuel storage capacity and spent fuel arising. The relevant IAEA activities in the area of spent fuel management are summarised. (author)

  4. Spent fuel management in Japan

    International Nuclear Information System (INIS)

    Shirahashi, K.; Maeda, M.; Nakai, T.

    1996-01-01

    Japan has scarce energy resources and depends on foreign resources for 84% of its energy needs. Therefore, Japan has made efforts to utilize nuclear power as a key energy source since mid-1950's. Today, the nuclear energy produced from 49 nuclear power plants is responsible for about 31% of Japan's total electricity supply. The cumulative amount of spent fuel generated as of March 1995 was about 11,600 Mg U. Japan's policy of spent fuel management is to reprocess spent nuclear fuel and recycle recovered plutonium and uranium as nuclear fuel. The Tokai reprocessing plant continues stable operation keeping the annual treatment capacity or around 90 Mg U. A commercial reprocessing plant is under construction at Rokkasho, northern part of Japan. Although FBR is the principal reactor to use plutonium, LWR will be a major power source for some time and recycling of the fuel in LWRs will be prompted. (author). 3 figs

  5. Spent fuel management in Canada

    International Nuclear Information System (INIS)

    Khan, A.; Pattantyus, P.

    1999-01-01

    The current status of the Canadian spent fuel storage is presented. This includes wet and dry interim storage. Extension of wet interim storage facilities is nor planned, as dry technologies have found wide acceptance. The Canadian nuclear program is sustained by commercial Ontario Hydro CANDU type reactors, since 1971, representing 13600 MW(e) of installed capacity, able to produce 9200 spent fuel bundles (1800 tU) every year, and Hydro Quebec and New Brunswick CANDU reactors each producing 685 MW(e) and about 100 tU of spent fuel annually. The implementation of various interim (wt and dry) storage technologies resulted in simple, dense and low cost systems. Economical factors determined that the open cycle option be adopted for the CANDU type reactors rather that recycling the spent fuel. Research and development activities for immobilization and final disposal of nuclear waste are being undertaken in the Canadian Nuclear Fuel Waste Management Program

  6. A seismic analysis of Korean standard PWR fuels under transition core conditions

    International Nuclear Information System (INIS)

    Kim, Hyeong Koo; Park, Nam Kyu; Jang, Young Ki; Kim, Jae Ik; Kim, Kyu Tae

    2005-01-01

    The PLUS7 fuel is developed to achieve higher thermal performance, burnup and more safety margin than the conventional fuel used in the Korean Standard Nuclear Plants (KSNPs) and to sustain structural integrity under increased seismic requirement in Korea. In this study, a series of seismic analysis have been performed in order to evaluate the structural integrity of fuel assemblies associated with seismic loads in the KSNPs under transition core conditions replacing the Guardian fuel, which is a resident fuel in the KSNP reactors, with the PLUS7 fuel. For the analysis, transition core seismic models have been developed, based on the possible fuel loading patterns. And the maximum impact forces on the spacer grid and various stresses acting on the fuel components have been evaluated and compared with the through-grid strength of spacer grids and the stress criteria specified in the ASME code for each fuel component, respectively. Then three noticeable parameters regarding as important parameters governing fuel assembly dynamic behavior are evaluated to clarify their effects on the fuel impact and stress response. As a result of the study, it has been confirmed that both the PLUS7 and the Guardian fuel sustain their structural integrity under the transition core condition. And when the damping ratio is constant, increasing the natural frequency of fuel assembly results in a decrease in impact force. The fuel assembly flexural stiffness has an effect increasing the stress of fuel assembly, but not the impact force. And the spacer grid stiffness is directly related with the impact force response. (author)

  7. PWR Core II blanket fuel disposition recommendation of storage option study

    International Nuclear Information System (INIS)

    Dana, C.M.

    1995-01-01

    After review of the options available for current storage of T Plant Fuel the recommended option is wet storage without the use of chillers. A test has been completed that verifies the maximum temperature reached is below the industrial standard for storage of spent fuel. This option will be the least costly and still maintain the fuel in a safe environment. The options that were evaluated included dry storage with and without chillers, and wet storage with and without chillers. Due to the low decay heat of the Shippingport Core II Blanket fuel assemblies the fuel pool temperature will not exceed 100 deg. F

  8. Commercializing fuel cells: managing risks

    Science.gov (United States)

    Bos, Peter B.

    Commercialization of fuel cells, like any other product, entails both financial and technical risks. Most of the fuel cell literature has focussed upon technical risks, however, the most significant risks during commercialization may well be associated with the financial funding requirements of this process. Successful commercialization requires an integrated management of these risks. Like any developing technology, fuel cells face the typical 'Catch-22' of commercialization: "to enter the market, the production costs must come down, however, to lower these costs, the cumulative production must be greatly increased, i.e. significant market penetration must occur". Unless explicit steps are taken to address this dilemma, fuel cell commercialization will remain slow and require large subsidies for market entry. To successfully address this commercialization dilemma, it is necessary to follow a market-driven commercialization strategy that identifies high-value entry markets while minimizing the financial and technical risks of market entry. The financial and technical risks of fuel cell commercialization are minimized, both for vendors and end-users, with the initial market entry of small-scale systems into high-value stationary applications. Small-scale systems, in the order of 1-40 kW, benefit from economies of production — as opposed to economies to scale — to attain rapid cost reductions from production learning and continuous technological innovation. These capital costs reductions will accelerate their commercialization through market pull as the fuel cell systems become progressively more viable, starting with various high-value stationary and, eventually, for high-volume mobile applications. To facilitate market penetration via market pull, fuel cell systems must meet market-derived economic and technical specifications and be compatible with existing market and fuels infrastructures. Compatibility with the fuels infrastructure is facilitated by a

  9. French analytic experiment on the high specific burnup of PWR fuels in normal conditions

    International Nuclear Information System (INIS)

    Bruet, M.; Atabek, R.; Houdaille, B.; Baron, D.

    1982-04-01

    Hydrostatic density determinations made on UO 2 pellets of different kinds irradiated in conditions representative of PWR conditions enable the internal swelling rate of the UO 2 to be ascertained. A mean value of 0.8% per 10 4 MWdt -1 (u) up to a specific burnup of 45000 MWdt -1 (u) may be deduced from this experimental basis. These results agree well with those obtained in the TANGO experiments in which UO 2 balls were irradiated in quasi isothermal conditions and without stress. Further, the open porosity of oxide closes progressively and the change in the total porosity is thus very limited (under 1% at 45000 MWdt -1 (u)). With respect to the swelling of the pellets the rise in the specific burnup would not appear therefore to be a problem. The behaviour of recrystallized zircaloy 4 claddings remains satisfactory with respect to creep and growth during irradiation [fr

  10. Aging management of reactor internals and license renewal of US PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Tang, H. T. [Electric Power Research Institute, EPRI, 3420 Hillview Avenue, Palo Alto, California 94304 (United States)

    2006-09-15

    Age-related degradation mechanisms of key components are subject to aging management review by utilities considering plant license renewal. The management of aging effects in PWR internals must be demonstrated as specified in the US NRC Standard Review. The US NRC staff has also issued a Generic Aging Lessons Learned (GALL) report that documents the staff's basis for determining when existing generic programs are adequate to manage aging without change and when existing generic programs should be augmented for license renewal. The EPRI Materials Reliability Program (MRP) has been conducting studies to develop technical bases and guidelines to support aging management of PWR internals, with a particular attention to utility License Renewal commitments. The strategic approach taken by the MRP includes: developing an overall aging management framework, defining degradation mechanism screening values, categorizing and ranking internals components based on screening, performing functionality analyses and safety evaluation, and developing inspection and evaluation guidelines associated with each category of components. Screening criteria are developed for the following potential internals degradation mechanisms: - Stress Corrosion Cracking [Excluding Irradiation Effects]; - Irradiation-Assisted Stress Corrosion Cracking; - Thermal Aging Embrittlement; - Irradiation Embrittlement; - Void Swelling; - Stress Relaxation and Creep [Irradiation-enhanced]; - Wear; - Fatigue. The ranking and categorization calls to bin internals components into four categories: - Category A: component items for which aging degradation significance is minimal and aging effects are below the screening criteria; - Category C: 'lead' component items for which aging degradation significance is high or moderate and aging effects are above screening levels; - Category B: component items above screening levels but are not 'lead' component items and aging degradation significance

  11. Spent Nuclear Fuel project, project management plan

    International Nuclear Information System (INIS)

    Fuquay, B.J.

    1995-01-01

    The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

  12. Experimentation, modelling and simulation of water droplets impact on ballooned sheath of PWR core fuel assemblies in a LOCA situation

    International Nuclear Information System (INIS)

    Lelong, Franck

    2010-01-01

    In a pressurized water reactor (PWR), during a Loss Of Coolant Accident (LOCA), liquid water evaporates and the fuel assemblies are not cooled anymore; as a consequence, the temperature rises to such an extent that some parts of the fuel assemblies can be deformed resulting in 'ballooned regions'. When reflooding occurs, the cooling of these partially blocked parts of the fuel assemblies will depend on the coolant flow that is a mixture of overheated vapour and under-saturated droplets. The aim of this thesis is to study the heat transfer between droplets and hot walls of the fuel rods. In this purpose, an experimental device has been designed in accordance with droplets and wall features (droplet velocity and diameter, wall temperature) representative of LOCA conditions. The cooling of a hot Nickel disk, previously heated by induction, is cooled down by a stream of monodispersed droplet. The rear face temperature profiles are measured by infrared thermography. Then, the estimation of wall heat flux is performed by an inverse conduction technique from these infrared images. The effect of droplet dynamical properties (diameter, velocity) on the heat flux is studied. These experimental data allow us to validate an analytical model of heat exchange between droplet and hot slab. This model is based on combined dynamical and thermal considerations. On the one hand, the droplet dynamics is considered through a spring analogy in order to evaluate the evolution of droplet features such as the spreading diameter when the droplet is squeezed over the hot surface. On the other hand, thermal parameters, such as the thickness of the vapour cushion beneath the droplet, are determined from an energy balance. In the short term, this model will be integrated in a CFD code (named NEPTUNE-CFD) to simulate the cooling of a reactor core during a LOCA, taking into account the droplet/wall heat exchange. (author)

  13. Generalized Thermohydraulics Module GENFLO for Combining With the PWR Core Melting Model, BWR Recriticality Neutronics Model and Fuel Performance Model

    International Nuclear Information System (INIS)

    Miettinen, Jaakko; Hamalainen, Anitta; Pekkarinen, Esko

    2002-01-01

    Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined codes. The basic field equations in GENFLO are for the phase mass, the mixture momentum and phase energy conservation equations. The phase separation is solved with the drift flux model. The basic variables to be solved are the pressure, void fraction, mixture velocity, gas enthalpy, liquid enthalpy, and concentration of non-condensable gas fractions. The validation of the thermohydraulic solution alone includes large break LOCA reflooding experiments and in specific for the severe accident conditions QUENCH tests. In the recriticality analysis the core neutronics is simulated with a two-dimensional transient neutronics code TWODIM. The recriticality with one rapid prompt peak is expected during a severe accident scenario, where the control rods have been melted and ECCS reflooding is started after the depressurization. The GENFLO module simulates the BWR thermohydraulics in this application. The core melting module has been developed for the real time operator training by using the APROS engineering simulators. The core heatup, oxidation, metal and fuel pellet relocation and corium pool formation into the lower plenum are calculated. In this application the GENFLO model simulates the PWR vessel thermohydraulics. In the fuel performance analysis the fuel rod transient behavior is simulated with the FRAPTRAN code. GENFLO simulates the subchannel around a single fuel rod and delivers the heat transfer on the cladding surface for the FRAPTRAN. The transient boundary conditions for the subchannel are transmitted from the system code for operational transient, loss of coolant accidents and

  14. Analysis of burnup of Angra 2 PWR nuclear with addition of thorium dioxide fuel using ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    Goncalves, Isadora C.; Wichrowski, Caio C.; Oliveira, Claudio L. de; Vellozo, Sergio O.; Baptista, Camila O., E-mail: isadora.goncalves@ime.eb.br, E-mail: wichrowski@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). Secao de Engenharia Nuclear

    2017-11-01

    It is known that isotope {sup 232}thorium is a fertile nuclide with the ability to convert into {sup 233}uranium, a potentially fissile isotope, after absorbing a neutron. As there is a large stock of available thorium in the world, this element shows great promise in mitigate the world energy crisis, more particularly in the problem of uranium scarcity, besides being an alternative nuclear fuel for those currently used in reactors, and yet presenting advantages as an option for the non-proliferation movement, among others. In this study, the analysis of the remaining nuclides of burnup was carried out for the core configuration of a PWR (pressurized water reactor) reactor, specifically the Angra 2 reactor, using only uranium dioxide, its current configuration, and in different configurations including a mixed oxide of uranium and thorium in three concentrations, allowing a preliminary assessment of the feasibility of the modification of the fuel, the resulting production of {sup 233}uranium, the emergence of {sup 231}protactinium (an isotope that only occurs as a fission product of {sup 232}Th) resulting from burning. The study was carried out using data obtained from FSAR (Final Safety Analysis Report) of Angra 2, using the SCALE 6.1, a modeling and simulation nuclear code, especially its ORIGEN-ARP module, which analyzes the depletion of isotopes presents in a reactor. (author)

  15. Preliminary performance analysis of exponential experimental system for the determination of neutron effective multiplication factor of PWR spent fuel

    International Nuclear Information System (INIS)

    Shin, Heesung; Lee, Sang-Yun; Ro, Seung-Gy; Seo, Gi-Seok; Kim, Ho-Dong

    2002-01-01

    An exponential experiment system which is composed of neutron detector, signal analysis system and neutron source, 10 mCi Cf-252 has been installed in the storage pool of PIEF at KAERI in order to experimentally determining neutron effective multiplication factors of PWR spent fuel assemblies. Preliminary functional characteristic tests of the experimental system are performed for C15, J14 and J44 assemblies loaded in the pool. As a result of preliminary tests, the average neutron counts obtained for 3 minutes in the plateau of the C15, J14 and J44 assemblies are about 1900, 3800 and 3200, respectively. A dip of the neutron flux density distribution is noticed in the spacer grid position. Neutron counts at those positions appear to be reduced to about 70 % in comparison to the fuel position. The measured axial neutron distribution shapes are compared with the result for the P14 assembly and Cs-137 gamma scanning data performed in KAERI. It is revealed that the spacer grid position measured is consistent with the design specifications within a 2.3 % error. The exponential decay constants for the C15 assembly were determined to be 0.152 and 0.165 for detector and source scanning, respectively. (author)

  16. In-plant test and evaluation of the neutron collar for verification of PWR fuel assemblies at Resende, Brazil

    International Nuclear Information System (INIS)

    Menlove, H.O.; Marzo, M.A.S.; de Almeida, S.G.; de Almeida, M.C.; Moitta, L.P.M.; Conti, L.F.; de Paiva, J.R.T.

    1985-11-01

    The neutron-coincidence collar has been evaluated for the measurement of pressurized-water reactor (PWR) fuel assemblies at the Fabrica de Elementos Combustiveis plant in Resende, Brazil. This evaluation was part of the cooperative-bilateral-safeguards technical-exchange program between the United States and Brazil. The neutron collar measures the 235 U content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The 238 U content is measured in the passive mode without the AmLi neutron-interrogation source. The extended evaluation took place over a period of 6 months with both scanning and single-zone measurements. The results of the tests gave a coincidence-response standard deviation of 0.7% (sigma = 1.49% for mass) for the active case and 2.5% for the passive case in 1000-s measurement times. The length measurement in the scanning mode was accurate to 0.77%. The accuracies of different calibration methods were evaluated and compared

  17. Analysis of burnup of Angra 2 PWR nuclear with addition of thorium dioxide fuel using ORIGEN-ARP

    International Nuclear Information System (INIS)

    Goncalves, Isadora C.; Wichrowski, Caio C.; Oliveira, Claudio L. de; Vellozo, Sergio O.; Baptista, Camila O.

    2017-01-01

    It is known that isotope "2"3"2thorium is a fertile nuclide with the ability to convert into "2"3"3uranium, a potentially fissile isotope, after absorbing a neutron. As there is a large stock of available thorium in the world, this element shows great promise in mitigate the world energy crisis, more particularly in the problem of uranium scarcity, besides being an alternative nuclear fuel for those currently used in reactors, and yet presenting advantages as an option for the non-proliferation movement, among others. In this study, the analysis of the remaining nuclides of burnup was carried out for the core configuration of a PWR (pressurized water reactor) reactor, specifically the Angra 2 reactor, using only uranium dioxide, its current configuration, and in different configurations including a mixed oxide of uranium and thorium in three concentrations, allowing a preliminary assessment of the feasibility of the modification of the fuel, the resulting production of "2"3"3uranium, the emergence of "2"3"1protactinium (an isotope that only occurs as a fission product of "2"3"2Th) resulting from burning. The study was carried out using data obtained from FSAR (Final Safety Analysis Report) of Angra 2, using the SCALE 6.1, a modeling and simulation nuclear code, especially its ORIGEN-ARP module, which analyzes the depletion of isotopes presents in a reactor. (author)

  18. CANDU-PHW fuel management

    International Nuclear Information System (INIS)

    Frescura, G.M.; Wight, A.L.

    1982-01-01

    This report covers the material presented in a series of six lectures at the Winter College on Nuclear Physics and Reactors held at the International Centre for Theoretical Physics in Trieste, Italy, Jan 22 - March 28, 1980. The report deals with fuel management in natural uranium fuelled CANDU-PHW reactors. Assuming that the reader has a basic knowledge of CANDU core physics, some of the reactor systems which are more closely related to fuelling are described. This is followed by a discussion of the methods used to calculate the power distribution and perform fuel management analyses for the equilibrium core. A brief description of some computer codes used in fuel management is given, together with an overview of the calculations required to provide parameters for core design and support the accident analysis. Fuel scheduling during approach to equilibrium and equilibrium is discussed. Fuel management during actual reactor operation is discussed with a review of the operating experience for some of the Ontario Hydro CANDU reactors. (author)

  19. Sustainomics of the AMBIDEXTER-NEC Fuel Cycle and Management

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Kee; Lee, Young Joon; Ham, Tae Kyu; Seo, Myung Hwan; Hong, Sung Taek; Kwon, Tae An [Ajou University, Suwon (Korea, Republic of)

    2009-05-15

    Energy issues these days become planetary concerns, recognized as the major driver for the resiliency of the earth in the sustainomics framework of the society, economy and environment axes. In the circumstances, in order for the nuclear to take advantage of its GHG-free nature, criticisms associated with the fuel cycle should be defied. As long as the uranium fuel cycle persists, problems bearing on the HLW management and the proliferation prevention could be neither completely decoupled nor independently resolved. Geopolitics around the Korean peninsula makes them be more complicated. Reference of the AMBIDEXTER fuel cycle relies on the DUPIC technology. Combined with fluoride volatility process, desired quantity of uranium contents in the PWR spent fuel powder could be removed. Then, the reactor system runs with the fluorides salt of this uranium-reduced DUPIC fuel material. Surplus uranium from the AMBIDEXTER-DUPIC1 processes should satisfy the LLW classification criteria. So far, the sustainomics goal of the AMBIDEXTER fuel cycle focuses on generating energy from the HLW, meanwhile, converting into LLW without jeopardizing proliferation transparency.

  20. Recent advances in PWR fuel design and performance experience at ABB-CENF

    International Nuclear Information System (INIS)

    Corsetti, Lawrence V.

    2004-01-01

    Utilities in the United States continue to move towards longer cycles and higher burnups to improve fuel cycle economics. This has placed increased demands for improved burnable absorber concepts. Zircaloy-4 corrosion behavior remains a high burnup performance concern. Over the past several years there has also been increasing utility interest in fuel reliability improvements. The development and application of erbia as a burnable absorber mixed directly with urania fuel will be discussed. Debris resistant fuel assembly designs and operating experience are reviewed. Oxide thickness measurements showing the improved corrosion resistance of optimized, low-tin Zircaloy-4 are presented. (author)

  1. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    International Nuclear Information System (INIS)

    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong

    2001-03-01

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3·4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident

  2. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    Energy Technology Data Exchange (ETDEWEB)

    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong [Seoul Nat' l Univ., Seoul (Korea, Republic of)

    2001-03-15

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3{center_dot}4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident.

  3. Spent fuel management and closed nuclear fuel cycle

    International Nuclear Information System (INIS)

    Kudryavtsev, E.G.

    2012-01-01

    Strategic objectives set by Rosatom Corporation in the field of spent fuel management are given. By 2030, Russia is to create technological infrastructure for innovative nuclear energy development, including complete closure of the nuclear fuel cycle. A target model of the spent NPP nuclear fuel management system until 2030 is analyzed. The schedule for key stages of putting in place the infrastructure for spent NPP fuel management is given. The financial aspect of the problem is also discussed [ru

  4. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  5. Development of a computer program for the cost analysis of spent fuel management

    International Nuclear Information System (INIS)

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

    2009-01-01

    So far, a substantial amount of spent fuels have been generated from the PWR and CANDU reactors. They are being temporarily stored at the nuclear power plant sites. It is expected that the temporary storage facility will be full of spent fuels by around 2016. The government plans to solve the problem by constructing an interim storage facility soon. The radioactive management act was enacted in 2008 to manage the spent fuels safety in Korea. According to the act, the radioactive waste management fund which will be used for the transportation, interim storage, and the final disposal of spent fuels has been established. The cost for the management of spent fuels is surprisingly high and could include a lot of uncertainty. KAERI and Kyunghee University have developed cost estimation tools to evaluate the cost for a spent fuel management based on an engineering design and calculation. It is not easy to develop a tool for a cost estimation under the situation that the national policy on a spent fuel management has not yet been fixed at all. Thus, the current version of the computer program is based on the current conceptual design of each management system. The main purpose of this paper is to introduce the computer program developed for the cost analysis of a spent fuel management. In order to show the application of the program, a spent fuel management scenario is prepared, and the cost for the scenario is estimated

  6. HFIR spent fuel management alternatives

    International Nuclear Information System (INIS)

    Begovich, J.M.; Green, V.M.; Shappert, L.B.; Lotts, A.L.

    1992-01-01

    The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems' Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere

  7. Probabilistic analysis of fuel pin behaviour during an eventual loss of coolant in PWR reactors

    International Nuclear Information System (INIS)

    1981-02-01

    Brief description of the development of the coolant loss incident in a pressurized water reactor and analysis of its significance for the behaviour of the fuel rods. Description of a probalistic method for estimating the effects of the accident on the fuel rods and results obtained [fr

  8. Specification of PWR UO2 pellet design parameters with the fuel performance code FRAPCON-1

    International Nuclear Information System (INIS)

    Silva, A.T.; Marra Neto, A.

    1988-08-01

    UO 2 pellet design parameters are analysed to verify their influence in the fuel basic properties and in its performance under irradiation in pressurized water reactors. Three groups of parameters are discussed: 1) content of fissionable and impurity materials; 2) stoichiometry; 3) density pore morpholoy, and microstructure. A methodology is applied with the fuel performance program FRAPCON-1 to specify these parameters. (author [pt

  9. PIN99W, Modelling of VVER and PWR Fuel Rod Thermomechanical Behaviour

    International Nuclear Information System (INIS)

    Valach, M.; Strizhov, P.; Svoboda, R.

    2000-01-01

    1 - Description of program or function: The Code is developed to describe fuel rod thermomechanical behaviour in operational conditions. The main goal of this code is to calculate fuel temperature, gap conductivity, fission gas release and inner gas pressure. 2 - Methods: - fuel rod temperature response is solved by using one-dimensional finite element method combined with weighted residuals method; - the code involves models describing physical phenomena typical for the fuel irradiated in Light Water Power Reactors (densification, restructuring, fission gas release, swelling and relocation) ; - this code is updated and improves PIN-micro code. 3 - Restrictions on the complexity of the problem: - simplified mechanistic solution; - only steady-state solution; - no cladding failure criterion; - no model for axial fuel-cladding interaction

  10. New technical knowledge to be implemented to the revision of rules on pipe wall thinning management for PWR Plants. 2006 edition

    International Nuclear Information System (INIS)

    Hirai, Junya; Amano, Yoichi; Nakamura, Takao

    2013-01-01

    Rules for PWR plant pipe wall thinning management were formulated by the Japan Society of Mechanical Engineers in 2006. Since then thinning management of Japanese PWR plants has been carried out based on this rule. Pipe wall thinning phenomena to be dealt with in this rule have been identified in many piping components of power plants. New technical knowledge has been accumulated since the issuance of 2006 edition. We have formulated these knowledge and information about the thinning phenomena in PWR power plants. Given the history of application of this rule, we have to make our best effort to carry out a study of latest technical knowledge and implement them to the revision of rule and improve pipe wall thinning management. This paper summarizes the new technical knowledge and basis to be implemented to the revision of rules on pipe wall thinning management for PWR plants in Japan. (author)

  11. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports

  12. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

  13. Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors

    International Nuclear Information System (INIS)

    Raepsaet, C.; Bossis, Ph.; Hamon, D.; Bechade, J.L.; Brachet, J.C.

    2007-01-01

    Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

  14. Corrosion product balances for the Ringhals PWR plants based on extensive fuel crud and water chemistry measurements

    International Nuclear Information System (INIS)

    Lundgren, K.; Wikmark, G.; Bengtsson, B.

    2010-01-01

    The corrosion product balance in a PWR plant is of great importance for the fuel performance as well as for the radiation field buildup. This balance is of special concern in connection to steam generator replacement (SGR) and power uprate projects. The Ringhals PWRs are all of Westinghouse design. Two of the plants have performed Steam Generator Replacement (SGR) to I-690 SG tubes and such a replacement is being planned in the third and last unit in 2011. Two of the units are in different phases of power uprate projects. The plants are all on 10-14-months cycles operating with medium to high fuel duty. Water chemistry is controlled by a pH300 in the range ∼7.2 to 7.4 from beginning of cycle to end of cycle (BOC-EOC) in the units with new SGs while kept at a coordinated pH of 7.2 in the one still using I-600. The maximum Li content has recently been increased to about 4.5 to 5 ppm in all units. In order to be able to improve the assessment of corrosion product balances in the plants, comprehensive fuel crud measurements were performed in 2007. Improved integrated reactor water sampling techniques have also been introduced in order to make accurate mass balances possible. The corrosion products covered in the study are the main constituents, Ni, Fe and Cr in the primary circuit Inconel and stainless steel, together with Co. The activated corrosion products, Co-58, Co-60, Cr-51, Fe-59 and Mn-54, are all mainly produced through neutron irradiation of the covered corrosion products. The main results of the corrosion product balances are presented. Observed differences between the plants, indicating significant impact of pH control and SG tube materials, are presented and discussed. The importance of accurate sampling techniques is especially addressed in this paper. (author)

  15. Impact of fission gas on irradiated PWR fuel behaviour at extended burnup under RIA conditions

    International Nuclear Information System (INIS)

    Lemoine, F.; Schmitz, F.

    1996-01-01

    With the world-wide trend to increase the fuel burnup at discharge of the LWRs, the reliability of high burnup fuel must be proven, including its behaviour under energetic transient conditions, and in particular during RIAs. Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup. The potential for swelling and transient expansion work under rapid heating conditions characterizes the high burnup fuel behaviour by comparison to fresh fuel. This effect is resulting from the steadily increasing amount of gaseous and volatile fission products retained inside the fuel structure. An attempt is presented to quantify the gas behaviour which is motivated by the results from the global tests both in CABRI and in NSRR. A coherent understanding of specific results, either transient release or post transient residual retention has been reached. The early failure of REP Na1 with consideration given to the satisfactory behaviour of the father rod of the test pin at the end of the irradiation (under load follow conditions) is to be explained both by the transient loading from gas driven fuel swelling and from the reduced clad resistance due to hydriding. (R.P.)

  16. Sensitivity Analysis of Depletion Parameters for Heat Load Evaluation of PWR Spent Fuel Storage Pool

    International Nuclear Information System (INIS)

    Kim, In Young; Lee, Un Chul

    2011-01-01

    As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

  17. Spent fuel management in Canada

    International Nuclear Information System (INIS)

    Pattantyus, P.

    1998-01-01

    The current status of the Canadian Spent Fuel Management is described. This includes wet and dry interim storage, transportation issues and future plans regarding final disposal based on deep underground emplacement in stable granite rock. Extension of wet interim storage facilities is not planned, as dry storage technologies have found wide acceptance. (author)

  18. Fluid flows due to leakproofness defects of nuclear fuel cans: application to sipping in PWR

    International Nuclear Information System (INIS)

    Parrat, D.

    1983-10-01

    This thesis establishes the theoritical laws describing fluid and gas transfers through channels representative of nuclear fuel can defects. These laws, applied to craks on fuel cans (samples) and on new or spent experimental rods, allow to know their size. A computerized modelling has been developed to analyse more particularly the examination by ''quantitative permeation testing'', giving the equivalent size of a channel from the observed signal. The utilization of this process to non-tight fuel assemblies of pressurized water reactors leads to the establishment of a classification of defects according to their importance [fr

  19. 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.)

  20. Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term-Disposal Criticality Safety

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1999-01-01

    Utilization of burnup credit in criticality safety analysis for long-term disposal of spent nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile material that will be present in the repository. Burnup-credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents (in terms of criticality potential), followed by criticality calculations to assess the value of the effective neutron multiplication factor (k(sub)eff) for the a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models

  1. Microstructural characteristics of PWR [pressurized water reactor] spent fuel relative to its leaching behavior

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1986-01-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data

  2. Eddy current testing of PWR fuel pencils in the pool of the Osiris reactor

    International Nuclear Information System (INIS)

    Faure, M.; Marchand, L.

    1983-12-01

    A nondestructive testing bench is described. It is devoted to examination of high residual power fuel pencils without stress on the cladding nor interference with cooling. Guiding by fluid bearings decrease the background noise. Scanning speed is limited only by safety criteria and data acquisition configuration. Simultaneous control of various parameters is possible. Associated to an irradiation loop, loaded and unloaded in a reactor swinning pool, this bench can follow fuel pencil degradation after each irradiation cycle [fr

  3. The experimental reactor Osiris and the nuclear fuel technology for the P.W.R. reactors

    International Nuclear Information System (INIS)

    Lestiboudois, G.; Contenson, G. de; Genthon, J.P.; Molvault, M.; Roche, M.

    1977-01-01

    The possibility of employing research reactors to study and to improve the nuclear fuel of the power reactors is presented. Measurements of temperature, pressure, stresses, thermal balance, gamma spectrometry and neutron radiography, allow the study of fuel densification, the influence of the initial filling pressure on the fission gas release and the gadolinium efficiency evolution. The solutions of the problems of failed element detection, power increase, remote handling, are presented [fr

  4. Production of actinide isotopes in simulated PWR fuel and their influence on inherent neutron emission

    International Nuclear Information System (INIS)

    Bosler, G.E.; Phillips, J.R.; Wilson, W.B.; LaBauve, R.J.; England, T.R.

    1982-07-01

    This report describes calculations that examine the sensitivity of actinide isotopes to various reactor parameters. The impact of actinide isotope build-up, depletion, and decay on the neutron source rate in a spent-fuel assembly is determined, and correlations between neutron source rates and spent-fuel characteristics such as exposure, fissile content, and plutonium content are established. The application of calculations for evaluating experimental results is discussed

  5. Influence of fuel vibration on PWR neutron noise associated with core barrel motion

    International Nuclear Information System (INIS)

    Sweeney, F.J.; March-Leuba, J.

    1984-01-01

    Ex-core neutron detector noise has been utilized to monitor core support barrel (CSB) vibrations. In order to observe long-term changes, noise signals at Sequoyah-1 were monitored continuously during the whole first fuel cycle and part of the second cycle. Results suggest that neutron noise measurements performed infrequently may not provide adequate surveillance of the CSB because it may be difficult to separate noise amplitude changes due solely to CSB motion from changes caused by fuel motion and burnup

  6. Management of irradiated CANDU fuel

    International Nuclear Information System (INIS)

    Lupien, Mario

    1985-01-01

    The nuclear industry, like any other industrial activity, generates waste and, since these radioactive products are known to be hazardous both to man and his natural environment, they are subject to stringent controls. The irradiated fuel is also highly radioactive and remains so for thousands of years. It is estimated that by the year 2000, nuclear reactors in Canada alone will have produced some 50 Gg of radioactive fuel which is stored at the nuclear plant site itself. The nuclear industry plays a leading role in the research and development effort to find suitable waste-management methods. Its R and D programs cover many scientific fields, including chemistry, and therefore demand a considerable amount of coordination. The knowledge acquired in this multidisciplinary context should form a basis for solving many of today's industrial-waste problems. This paper describes the various stages in the long management process. In the medium term, the irradiated fuel will be stored in surface installations but the long-term solution proposed is to emplace the used fuel or the fuel recycle waste deep underground in a stable geologic formation

  7. The Gd-isotopic fuel for high burnup in PWR's

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Marcio Soares; Mattos, João Roberto L. de; Andrade, Edison Pereira de, E-mail: marciod@cdtn.br, E-mail: jrmattos@cdtn.br, E-mail: epa@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    Today, the discussion about the high burnup fuel is beyond the current fuel enrichment licensing and burnup limits. Licensing issues and material/design developments are again key features in further development of the LWR fuel design. Nevertheless, technological and economical solutions are already available or will be available in a short time. In order to prevent the growth of the technological gap, Brazil's nuclear sector needs to invest in the training of new human resources, in the access to international databases, and in the upgrading existing infrastructure. Experimental database and R&D infrastructure are essential components to support the autonomous development of Brazilian Nuclear Reactors, promoting the development of national technologies. The (U,Gd)O{sub 2} isotopic fuel proposed by the CDTN's staff solve two main issues in the high burnup fuel, which are (1) the peak of reactivity resulting from the Gd-157 fast burnup, and (2) the peak of temperature in the (U,Gd)O{sub 2} nuclear fuel resulting from detrimental effects in the thermal properties for gadolinia additions higher than 2%. A sustainable future can be envisaged for the nuclear energy. (author)

  8. 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)

  9. Development of Experimental Facilities for Advanced Spent Fuel Management Technology

    Energy Technology Data Exchange (ETDEWEB)

    You, G. S.; Jung, W. M.; Ku, J. H. [and others

    2004-07-01

    The advanced spent fuel management process(ACP), proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel, is under research and development. This technology convert spent fuels into pure metal-base uranium with removing the highly heat generating materials(Cs, Sr) efficiently and reducing of the decay heat, volume, and radioactivity from spent fuel by 1/4. In the next phase(2004{approx}2006), the demonstration of this technology will be carried out for verification of the ACP in a laboratory scale. For this demonstration, the hot cell facilities of {alpha}-{gamma} type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of {beta}-{gamma} type will be refurbished to minimize construction expenditures of hot cell facility. In this study, the design requirements are established, and the process detail work flow was analysed for the optimum arrangement to ensure effective process operation in hot cell. And also, the basic and detail design of hot cell facility and process, and safety analysis was performed to secure conservative safety of hot cell facility and process.

  10. CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety

    2011-07-01

    been developed for calculation and synchronization purposes. The data exchange is realized by means of the Parallel Virtual Machine (PVM) software package. In this contribution, steady-state and transient results of a quarter of PWR fuel assembly with cold water injection are presented and compared with obtained results from a RELAP5/PARCS v2.7 coupled calculation. A simplified model for the spacers has been included. A methodology has been introduced to take into account the pressure drop and the turbulence enhancement produced by the spacers. (author)

  11. CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

    International Nuclear Information System (INIS)

    Pena, C.; Pellacani, F.; Macian Juan, R.; Chiva, S.; Barrachina, T.; Miro, R.

    2011-01-01

    developed for calculation and synchronization purposes. The data exchange is realized by means of the Parallel Virtual Machine (PVM) software package. In this contribution, steady-state and transient results of a quarter of PWR fuel assembly with cold water injection are presented and compared with obtained results from a RELAP5/PARCS v2.7 coupled calculation. A simplified model for the spacers has been included. A methodology has been introduced to take into account the pressure drop and the turbulence enhancement produced by the spacers. (author)

  12. Spent fuel management in Spain

    International Nuclear Information System (INIS)

    Gago, J.A.; Gravalos, J.M.

    1996-01-01

    There are presently nine Light Water Reactors in operation, representing around a 34% of the overall electricity production. In the early years, a small amount of spent fuel was sent to be reprocessed, although this policy was cancelled in favor of the open cycle option. A state owned company, ENRESA, was created in 1984, which was given the mandate to manage all kinds of radioactive wastes generated in the country. Under the present scenario, a rough overall amount of 7000 tU of spent fuel will be produced during the lifetime of the plants, which will go into final disposal. (author)

  13. Development of unfolding method to obtain pin-wise source strength distribution from PWR spent fuel assembly measurement

    International Nuclear Information System (INIS)

    Sitompul, Yos Panagaman; Shin, Hee-Sung; Park, Se-Hwan; Oh, Jong Myeong; Seo, Hee; Kim, Ho Dong

    2013-01-01

    An unfolding method has been developed to obtain a pin-wise source strength distribution of a 14 × 14 pressurized water reactor (PWR) spent fuel assembly. Sixteen measured gamma dose rates at 16 control rod guide tubes of an assembly are unfolded to 179 pin-wise source strengths of the assembly. The method calculates and optimizes five coefficients of the quadratic fitting function for X-Y source strength distribution, iteratively. The pin-wise source strengths are obtained at the sixth iteration, with a maximum difference between two sequential iterations of about 0.2%. The relative distribution of pin-wise source strength from the unfolding is checked using a comparison with the design code (Westinghouse APA code). The result shows that the relative distribution from the unfolding and design code is consistent within a 5% difference. The absolute value of the pin-wise source strength is also checked by reproducing the dose rates at the measurement points. The result shows that the pin-wise source strengths from the unfolding reproduce the dose rates within a 2% difference. (author)

  14. A practical approach to burn-up credit use in package design approval for PWR uranium oxide spent fuel assemblies

    International Nuclear Information System (INIS)

    Kroger, H.; Reiche, I.

    2009-01-01

    TN International has applied for a license for the TN 24 E transport and storage cask with the German competent authority using a new Burn-up Credit (BUC) approach for PWR uranium oxide fuel assemblies based on actinides and six selected fission products. In order to enable the use of BUC for fission products, various experimental data have to be provided for the two important aspects of the criticality calculation. Firstly, post-irradiation examination (PIE) experiments for the verification of the calculated fission product concentrations have to be provided for each selected fission product. These data are then used to validate the depletion calculations. Secondly, experimental data for the criticality calculations in the form of critical benchmark experiments have to be provided. The submitted data will be investigated for their applicability to the TN 24 E transport and storage cask. Since the application is limited to six fission products only, the conservatism of the BUC approach can be further justified, as the reduction in reactivity from the remaining fission products (about 190) is not taken credit for. (authors)

  15. Assembly-level analysis of heterogeneous Th–Pu PWR fuel

    International Nuclear Information System (INIS)

    Zainuddin, Nurjuanis Zara; Parks, Geoffrey T.; Shwageraus, Eugene

    2017-01-01

    Highlights: • We directly compare homogeneous and heterogeneous Th–Pu fuel. • Examine whether there is an increase in Pu incineration in the latter. • Homogeneous fuel was able to achieve much higher Pu incineration. • In the heterogeneous case, U-233 breeding is faster (larger power fraction), thus decreasing incineration of Pu. - Abstract: This study compares homogeneous and heterogeneous thorium–plutonium (Th–Pu) fuel assemblies (with high Pu content – 20 wt%), and examines whether there is an increase in Pu incineration in the latter. A seed-blanket configuration based on the Radkowsky thorium reactor concept is used for the heterogeneous assembly. This separates the thorium blanket from the uranium seed, or in this case a plutonium seed. The seed supplies neutrons to the subcritical thorium blanket which encourages the in situ breeding and burning of "2"3"3U, allowing the fuel to stay critical for longer, extending burnup of the fuel. While past work on Th–Pu seed-blanket units shows superior Pu incineration compared to conventional U–Pu mixed oxide fuel, there is no literature to date that directly compares the performance of homogeneous and heterogeneous Th–Pu assembly configurations. Use of exactly the same fuel loading for both configurations allows the effects of spatial separation to be fully understood. It was found that the homogeneous fuel with and without burnable poisons was able to achieve much higher Pu incinerations than the heterogeneous fuel configurations, while still attaining a reasonably high discharge burnup. This is because in the heterogeneous cases, "2"3"3U breeding is faster, thereby contributing to a much larger fraction of total power produced by the assembly. In contrast, "2"3"3U build-up is slower in the homogeneous case and therefore Pu burning is greater. This "2"3"3U begins to contribute a significant fraction of power produced only towards the end of life, thus extending criticality, allowing more Pu to

  16. Response of unirradiated and irradiated PWR fuel rods tested under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Quapp, W.J.; Martinson, Z.R.; McCardell, R.K.; Mehner, A.S.

    1978-01-01

    This report summarizes the results from the single-rod power-cooling-mismatch (PCM) and irradiation effects (IE) tests conducted to date in the Power Burst Facility (PBF) at the U.S. DOE Idaho National Engineering Laboratory. This work was performed for the U.S. NRC under contact to the Department of Energy. These tests are part of the NRC Fuel Behavior Program, which is designed to provide data for the development and verification of analytical fuel behavior models that are used to predict fuel response to abnormal or postulated accident conditions in commercial LWRs. The mechanical, chemical and thermal response of both previously unirradiated and previously irradiated LWR-type fuel rods tested under power-cooling-mismatch condition is discussed. A brief description of the test designs is presented. The results of the PCM thermal-hydraulic studies are summarized. Primary emphasis is placed on the behavior of the fuel and cladding during and after stable film boiling. (orig.) [de

  17. Development of a Computer Program for the Analysis Logistics of PWR Spent Fuels

    International Nuclear Information System (INIS)

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

    2008-01-01

    It is expected that the temporary storage facilities at the nuclear power plants will be full of the spent fuels within 10 years. Provided that a centralized interim storage facility is constructed along the coast of the Korean peninsula to solve this problem, a substantial amount of spent fuels should be transported by sea or by land every year. In this paper we developed a computer program for the analysis of transportation logistics of the spent fuels from 4 different nuclear power plant sites to the hypothetical centralized interim storage facility and the final repository. Mass balance equations were used to analyze the logistics between the nuclear power plants and the interim storage facility. To this end a computer program, CASK, was developed by using the VISUAL BASIC language. The annual transportation rates of spent fuels from the four nuclear power plant sites were determined by using the CASK program. The parameter study with the program illustrated the easiness of logistics analysis. The program could be used for the cost analysis of the spent fuel transportation as well.

  18. Conservative performance analysis of a PWR nuclear fuel rod using the FRAPCON code

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Fabio Branco Vaz de; Sabundjian, Gaiane, E-mail: fabio@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    In this paper, some of the preliminary results of the sensitivity and conservative analysis of a hypothetical pressurized water reactor fuel rod are presented, using the FRAPCON code as a basic and preparation tool for the future transient analysis, which will be carried out by the FRAPTRAN code. Emphasis is given to the evaluation of the cladding behavior, since it is one of the critical containment barriers of the fission products, generated during fuel irradiation. Sensitivity analyses were performed by the variation of the values of some parameters, which were mainly related with thermal cycle conditions, and taking into account an intermediate value between the realistic and conservative conditions for the linear heat generation rate parameter, given in literature. Time lengths were taken from typical nuclear power plant operational cycle, adjusted to the obtention of a chosen burnup. Curves of fuel and cladding temperatures, and also for their mechanical and oxidation behavior, as a function of the reactor operation's time, are presented for each one of the nodes considered, over the nuclear fuel rod. Analyzing the curves, it was possible to observe the influence of the thermal cycle on the fuel rod performance, in this preliminary step for the accident/transient analysis. (author)

  19. Power cycling and ramp test in R-2 and Mihama Unit 2 for MHI PWR fuels

    International Nuclear Information System (INIS)

    Baba, T.; Takahashi, T.; Kubo, H.; Fujiwara, Y.; Kondo, Y.

    1983-01-01

    Up to the present time, Mitsubishi has manufactured approximately 3000 fuel assemblies for Japanese PWRs, of which performance in reactors is satisfactory under base load operation. For the forthcoming load following age in Japan, expected in mid eighties, Mitsubishi is performing various R and D programs, so that load following operation can be smoothly introduced with current good performance maintained. R and D programs consist of two phases. One is the verification and demonstration of power ramping and cycling capability of the current design fuels, and the other is the development of remedy fuels with more operational margin. This paper describes the recent results obtained for the former phase, especially for the following two programs: (1) Power cycling and ramp test in R-2; (2) Power ramp demonstration (PRD) in Mihama Unit 2 (PRD-1). PIE works for power cycling and ramp test rods have been almost completed. The second PRD will be performed early in 1983

  20. Artificial neural networks for spatial distribution of fuel assemblies in reload of PWR reactors

    International Nuclear Information System (INIS)

    Oliveira, Edyene; Castro, Victor F.; Velásquez, Carlos E.; Pereira, Claubia

    2017-01-01

    An artificial neural network methodology is being developed in order to find an optimum spatial distribution of the fuel assemblies in a nuclear reactor core during reload. The main bounding parameter of the modelling was the neutron multiplication factor, k ef f . The characteristics of the network are defined by the nuclear parameters: cycle, burnup, enrichment, fuel type, and average power peak of each element. These parameters were obtained by the ORNL nuclear code package SCALE6.0. As for the artificial neural network, the ANN Feedforward Multi L ayer P erceptron with various layers and neurons were constructed. Three algorithms were used and tested: LM (Levenberg-Marquardt), SCG (Scaled Conjugate Gradient) and BayR (Bayesian Regularization). Artificial neural network have implemented using MATLAB 2015a version. As preliminary results, the spatial distribution of the fuel assemblies in the core using a neural network was slightly better than the standard core. (author)

  1. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations

    International Nuclear Information System (INIS)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-01-01

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k eff of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data

  2. Project management of the build of the shore test facility for the prototype of PWR II

    International Nuclear Information System (INIS)

    Clarkson, D.T.

    1987-01-01

    The PWR II is a new design of nuclear steam raising plant for the Royal Navy's submarines. It features improved engineering for safety, increased power, increased shock resistance, reduced noise transmission to sea and reduced manning requirement. It is to be tested in a new prototype testing facility, the Shore Test Facility, which is a section of submarine hull containing a prototype of the nuclear steam raising plant and its support system. It is installed at the Vulcan Naval Reactor Test establishment at Dounreay in Scotland. The function of the establishment is to test new designs of core and reactor plant, validate the mathematical models used in their design, develop improved methods of operation and maintenance of the plant and test new items of equipment. The Shore Test Facility was built in large sections at Barrow-in-Furness and transported to Scotland. The project management for the construction of the Shore Test Facility is explained. It involves personnel from the Royal Navy, and a large number of people working for the contractors involved in the buildings, transportation, operation and maintenance of the Facility. (U.K.)

  3. Surveillance of PLUS7TM fuel for PWR nuclear power plant

    International Nuclear Information System (INIS)

    Jang, Y. K.; Kim, J. I.; Shin, J. C.; Chung, J. G.; Chung, S. K.; Kim, M. S.; Lee, T. H.; Yoon, Y. B.; Kim, T. W.

    2012-01-01

    The surveillance program on the advanced nuclear fuel of PLUS TM developed for Optimized Power Reactors of 1000 MWe (OPR1000s) and Advanced Power Reactors of 1400 MWe (APR1400s) in Korea was completed in the early of 2011. This fuel had been jointly developed through the extensive out-of-pile tests with Westinghouse for three years since 1999. The irradiation tests for the in-reactor verification using four lead test assemblies (LTAs) had been started in Ulchin unit 3 in 2002. During the overhaul period after each irradiation test, the eight (8) burnup-dependent parameters were measured without disassembling using the precise measurement systems in pool-side. After three cycle irradiations, one test assembly was disassembled and the rod-wise inspection on twenty rods was performed. During this stage, five (5) parameters were measured and evaluated. Among these twenty rods, ten rods including skeleton were sent to hot-cell test facility for further detailed examination and are currently being examined. After in-reactor verifications during two cycles, this fuel was commercially supplied to eight (8) OPR1000s sequentially. Currently all eight (8) OPR1000s were replaced with this fuel. In addition, this fuel is going to be supplied to four (4) APR1400s being constructed in Braka, UAE as well as four(4) OPR1000s and four(4) APR1400s being constructed in Korea. In the meanwhile, the surveillance program for the commercially supplied fuel has been launched to confirm growth, creep, corrosion and deformation, etc. obtained during LTA irradiation. Four (4) limiting fuel assemblies, that is, two (2) assemblies to be discharged after 2 cycle irradiations and the other two (2) after 3 cycle irradiations were selected for this surveillance program. Irradiation data of commercially supplied fuels are compared and confirmed to LTA irradiation performance behaviors on this paper. Among the eight (8) burnup-dependent parameters, the interesting ones were irradiation

  4. Mechanical characteristic evaluation of the mid grid spring in PWR fuel assembly

    International Nuclear Information System (INIS)

    Eum, K. B.; Lee, S. H.; Jeon, S. Y.; Kweon, Y. B.; Jeon, K. R.

    2001-01-01

    The spring load-deflection characteristic tests were performed for Westinghouse type 17x17 and 14x14 fuel assembly mid grids to evaluate the mechanical characteristics of the springs. Six kinds of prototype mid grids manufactured by KNFC were tested and two kinds of test methods were used: block test and in-grid test. The test results showed that all tested mid grid springs satisfied the criteria required at the beginning of fuel assembly life. In addition, the variation of spring characteristics resulting from the difference in the mechanical properties of spring material and spring shapes was investigated. And the validity of the test methods was discussed

  5. Seismic testing of the base-isolated PWR spent-fuel storage rack

    International Nuclear Information System (INIS)

    Fujita, Katsuhisa; Tanaka, Mamoru; Nakamura, Masaaki; Tsujikura, Yonezo.

    1990-01-01

    The present paper aims to verify the seismic safety of the base-isolated spent-fuel storage rack. A series of seismic tests has been conducted using a three-dimensional shaking table. A sliding-type base-isolation system was employed for the prototype rack considering environmental conditions in an actual plant. A non linear seismic response analysis was also performed, and it is verified that the prototype of a base-isolated spent-fuel storage rack has a sufficient seismic safety margin for design seismic conditions from the viewpoint of seismic response. (author)

  6. A study of solute transport of radiolysis products in crud and its effects on crud growth on PWR fuel pin

    Energy Technology Data Exchange (ETDEWEB)

    Joe, Justin H. [BNF Consulting (United States); Kim, Seung Jun, E-mail: skim@lanl.gov [Mechanical and Thermal Engineering Group (AET-1), Los Alamos National Laboratory (United States); Jones, Barclay G. [Department of Nuclear Plasma Radiological Engineering, University of Illinois Urbana-Champaign (United States)

    2016-04-15

    Highlights: • We model a 3-D numerical solute transport within crud deposit on PWR fuel pin. • Source term effect from radiolysis yield and recombination is minimal. • Lower crud porosity leads substantially higher concentration of solutes. • Thicker crud deposit generates substantially higher concentration of solutes. • High concentration of radiolysis species (H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2}) can be directly related to corrosion issues on fuel cladding. - Abstract: This research examines the concentration of radiolysis species (H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2}) over the porous crud layer using a three dimensional time dependent solute transport model. A Monte Carlo random walk technique is adopted to simulate the transport behavior of the different species with various parametric studies of source term, crud thickness, and crud porosity. Particularly, this model employs a system of coupled mass transport and chemical interactions as the source term, which makes the problem non-linear. It is demonstrated that a negligible effect on radiolysis species concentrations change due to the consideration of source term. The crud thickness and porosity effect on the concentration distributions are notably observed. In general, higher concentration starts from the intersection of the heating surface with the chimney wall from the beginning and it reaches the equilibrium state within tens of seconds. The concentration profiles of the radiolysis species H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2} can be directly related to corrosion issues. The direct application of this study to nuclear engineering research is to aid in the design of reactors with higher performance without experiencing an Axial Offset Anomaly (AOA), an unexpected measured shift in axial power distribution from predicted values.

  7. A study of solute transport of radiolysis products in crud and its effects on crud growth on PWR fuel pin

    International Nuclear Information System (INIS)

    Joe, Justin H.; Kim, Seung Jun; Jones, Barclay G.

    2016-01-01

    Highlights: • We model a 3-D numerical solute transport within crud deposit on PWR fuel pin. • Source term effect from radiolysis yield and recombination is minimal. • Lower crud porosity leads substantially higher concentration of solutes. • Thicker crud deposit generates substantially higher concentration of solutes. • High concentration of radiolysis species (H 2 , O 2 , and H 2 O 2 ) can be directly related to corrosion issues on fuel cladding. - Abstract: This research examines the concentration of radiolysis species (H 2 , O 2 , and H 2 O 2 ) over the porous crud layer using a three dimensional time dependent solute transport model. A Monte Carlo random walk technique is adopted to simulate the transport behavior of the different species with various parametric studies of source term, crud thickness, and crud porosity. Particularly, this model employs a system of coupled mass transport and chemical interactions as the source term, which makes the problem non-linear. It is demonstrated that a negligible effect on radiolysis species concentrations change due to the consideration of source term. The crud thickness and porosity effect on the concentration distributions are notably observed. In general, higher concentration starts from the intersection of the heating surface with the chimney wall from the beginning and it reaches the equilibrium state within tens of seconds. The concentration profiles of the radiolysis species H 2 , O 2 , and H 2 O 2 can be directly related to corrosion issues. The direct application of this study to nuclear engineering research is to aid in the design of reactors with higher performance without experiencing an Axial Offset Anomaly (AOA), an unexpected measured shift in axial power distribution from predicted values.

  8. Assessment of management alternatives for LWR wastes. Volume 4. Description of a Belgian scenario for PWR waste

    International Nuclear Information System (INIS)

    Crustin, J.; Glibert, R.

    1993-01-01

    This report deals with the description of a management route for PWR waste relying to a certain extent on Belgian practices in this particular area. This description, which aims at providing input data for subsequent cost evaluation, includes all management steps which are usually implemented for solid, liquid and gaseous wastes from their production up to the interim storage of the final waste products. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for LWR waste based on economical and radiological criteria

  9. Assessment of management alternatives for LWR wastes. Volume 2. Description of a French scenario for PWR waste

    International Nuclear Information System (INIS)

    Saulieu, E. de; Chary, C.

    1993-01-01

    This report deals with the description of a management route for PWR waste relying to a certain extent on French practices in this particular area. This description, which aims at providing input data for subsequent cost evaluation, includes all management steps which are usually implemented for solid, liquid and gaseous wastes from their production up to the interim storage of the final waste products. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for LWR waste based on economical and radiological criteria

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  12. A study on the generation of radioactive corrosion product at PWR for extended fuel cycle

    International Nuclear Information System (INIS)

    Min Chul Song; Kun Jai Lee

    2001-01-01

    Current nuclear power plant operating practice is to extend the time between refueling from a 12 month operating cycle to an 18-24 month period. This current to longer fuel cycles has complicated the dilemma of finding optimum pH range for the primary coolant chemistry. The International Commission on Radiological Protection (ICRP) in ICRP publication No. 60 recommends optimization of operator radiation exposure (ORE) in nuclear power plants. CRUD formed in the plants is the major source of ORE and its transport mechanism is not understood. To analyze the generation of CRUD at the extended fuel cycle, the COTRAN code, which was developed at the Korea Advanced Institute of Science and Technology (KAIST), was used. It predicts that the activity of CRUD decreases as the pH of the coolant increases. For the same period of different fuel cycles, as the operating fuel cycle duration is increased, the generation of the CRUD increases. In this paper, enriched boric acid (40% enriched 10 B concentration) for reactivity control is adopted as the required chemical shim rather than natural boric acid. The effect of the enriched boric acid (EBA) is that the neutron absorption capability of the chemical shim is maintained while decreasing the required boron and lithium concentration in the reactor coolant system. By employing enriched boric acid, the amounts of CRUD generated are reduced, because the high pH-operating period is extended. From the waste generation point of view, more filters or ion exchangers to remove CRUD are required and the amounts of waste are increased at the extended fuel cycle. (author)

  13. Experience of Areva in fuel services for PWR and BWR; Experiencia de Areva en servicios de combustible para PWR y BWR

    Energy Technology Data Exchange (ETDEWEB)

    Morales, I.

    2015-07-01

    AREVA being an integrated supplier of fuel assemblies has included in its strategy to develop services and solutions to customers who desire to improve the performance and safety of their fuel. These services go beyond the simple 'after sale' services that can be expected from a fuel supplier: The portfolio of AREVA includes a wide variety of services, from scientific calculations to fuel handling services in a nuclear power plant. AREVA is committed to collaborate and to propose best-in-class solutions that really make the difference for the customer, based on 40 years of Fuel design and manufacturing experience. (Author)

  14. Validation of PWR core seismic models with shaking table tests on interacting scale 1 fuel assemblies

    International Nuclear Information System (INIS)

    Viallet, E.; Bolsee, G.; Ladouceur, B.; Goubin, T.; Rigaudeau, J.

    2003-01-01

    The fuel assembly mechanical strength must be justified with respect to the lateral loads under accident conditions, in particular seismic loads. This justification is performed by means of time-history analyses with dynamic models of an assembly row in the core, allowing for assembly deformations, impacts at grid locations and reactor coolant effects. Due to necessary simplifications, the models include 'equivalent' parameters adjusted with respect to dynamic characterisation tests of the fuel assemblies. Complementing such tests on isolated assemblies by an overall model validation with shaking table tests on interacting assemblies is obviously desirable. Seismic tests have been performed by French CEA (Commissariat a l'Energie Atomique) on a row of six full scale fuel assemblies, including two types of 17 x 17 12ft design. The row models are built according to the usual procedure, with preliminary characterisation tests performed on a single assembly. The test-calculation comparisons are made for two test configurations : in air and in water. The relatively large number of accelerograms (15, used for each configuration) is also favourable to significant comparisons. The results are presented for the impact forces at row ends, displacements at mid assembly, and also 'statistical' parameters. Despite a non-negligible scattering in the results obtained with different accelerograms, the calculations prove realistic, and the modelling process is validated with a good confidence level. This satisfactory validation allows to evaluate precisely the margins in the seismic design methodology of the fuel assemblies, and thus to confirm the safety of the plants in case of seismic event. (author)

  15. PHEBUS program: first results on PWR fuel behaviour in LOCA conditions

    International Nuclear Information System (INIS)

    Del Negro, R.; Reocreux, M.; Pelce, J.; Legrand, B.; Berna, P.

    1982-09-01

    In the first PHEBUS test with pressurized rods some rods burst and clad temperature reached 1100 0 C in the 25 rods bundle. There is now a lot of valuable experimental results and their analysis is in progress. The phase II on fuel behaviour in case of a large LOCA will start at the beginning of 83. The onset of the SFD program is foreseen to take place on the first months of 85

  16. Study of a brazilian cask and its installation for PWR spent nuclear fuel dry storage

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2009-01-01

    Spent nuclear fuel (SNF) is removed from the nuclear reactor after the depletion on efficiency in generating energy. After the withdrawal from the reactor core, the SNF is temporarily stored in pools at the same site of the reactor. At this time, the generated heat and the short and medium lived radioactive elements decay to levels that allow removing SNF from the pool and sending it to temporary dry storage. In that phase, the fuel needs to be safely and efficiently stored, and then, it can be retrieved in a future, or can be disposed as radioactive waste. The amount of spent fuel increases annually and, in the next years, will still increase more, because of the construction of new nuclear plants. Today, the number of new facilities back up to levels of the 1970's, since it is greater than the amount of decommissioning in old installations. As no final decision on the back-end of the nuclear fuel cycle is foreseen in the near future in Brazil, either to recover the SNF or to consider it as radioactive waste, this material has to be isolated in some type of storage model existing around the world. In the present study it is shown that dry SNF storage is the best option. A national cask model for SNF as well these casks storage installation are proposed. It is a multidisciplinary study in which the engineering conceptual task was developed and may be applied to national SNF removed from the Brazilian power reactors, to be safely stored for a long time until the Brazilian authorities will decide about the site for final disposal. (author)

  17. Artificial neural networks for spatial distribution of fuel assemblies in reload of PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Edyene; Castro, Victor F.; Velásquez, Carlos E.; Pereira, Claubia, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Programa de Pós-Graduação em Ciências e Técnicas Nucleares

    2017-07-01

    An artificial neural network methodology is being developed in order to find an optimum spatial distribution of the fuel assemblies in a nuclear reactor core during reload. The main bounding parameter of the modelling was the neutron multiplication factor, k{sub ef{sub f}}. The characteristics of the network are defined by the nuclear parameters: cycle, burnup, enrichment, fuel type, and average power peak of each element. These parameters were obtained by the ORNL nuclear code package SCALE6.0. As for the artificial neural network, the ANN Feedforward Multi{sub L}ayer{sub P}erceptron with various layers and neurons were constructed. Three algorithms were used and tested: LM (Levenberg-Marquardt), SCG (Scaled Conjugate Gradient) and BayR (Bayesian Regularization). Artificial neural network have implemented using MATLAB 2015a version. As preliminary results, the spatial distribution of the fuel assemblies in the core using a neural network was slightly better than the standard core. (author)

  18. Improved and consistent determination of the nuclear inventory of spent PWR-fuel on the basis of time-dependent cell-calculations with KORIGEN

    International Nuclear Information System (INIS)

    Fischer, U.; Wiese, H.W.

    1983-01-01

    For safe handling, processing and storage of spent nuclear fuel a reliable, experimentally validated method is needed to determine fuel and waste characteristics: composition, radioactivity, heat and radiation. For PWR's, a cell-burnup procedure has been developed which is able to calculate the inventory in consistency with cell geometry, initial enrichment, and reactor control. Routine calculations can be performed with KORIGEN using consistent cross-section sets - burnup-dependent and based on the latest Karlsruhe evaluations for actinides - which were calculated previously with the cell-burnup procedure. Extensive comparisons between calculations and experiments validate the presented procedure. For the use of the KORIGEN code the input description and sample problems are added. Improvements in the calculational method and in data are described, results from KORIGEN, ORIGEN and ORIGEN2 calculations are compared. Fuel and waste inventories are given for BIBLIS-type fuel of different burnup. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  20. 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

  1. 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

  2. Nuclear fuel management via fuel quality factor averaging

    International Nuclear Information System (INIS)

    Mingle, J.O.

    1978-01-01

    The numerical procedure of prime number averaging is applied to the fuel quality factor distribution of once and twice-burned fuel in order to evolve a fuel management scheme. The resulting fuel shuffling arrangement produces a near optimal flat power profile both under beginning-of-life and end-of-life conditions. The procedure is easily applied requiring only the solution of linear algebraic equations. (author)

  3. 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

  4. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant; Avaliacao da integridade de varetas combustiveis em reatores PWR a partir da analise espectrometrica da agua do primario

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Iara Arraes

    1999-02-15

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  5. Experimental study of the deformation of Zircaloy PWR fuel rod cladding under mainly convective cooling

    International Nuclear Information System (INIS)

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

    1982-01-01

    Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 915 degree C in flowing steam at atmospheric pressure. Internal test pressures were in the range 0.69 to 11.0 MPa. The length of cladding strained 33 percent or more was greatest (about 20 times the original diameter) when the initial pressure was 1.38/plus or minus/0.17MPa. This results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilizing the deformation or partial superplastic deformation, or both. For adjacent rods in a fuel assembly not to touch at any temperature, the pressure would have to be less than about 1 MPa. These results are compared with those form multirod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behavior of fuel elements in a loss-of-coolant accident are outlined. 37 refs

  6. Experimental study of the deformation of Zircaloy PWR fuel rod cladding under mainly convective cooling

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-01-01

    Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 915 degree C in flowing steam at atmospheric pressure. Internal test pressures were in the range 0.69 to 11.0 MPa. The length of cladding strained 33 percent or more was greatest (about 20 times the original diameter) when the initial pressure was 1.38/plus or minus/0.17MPa. This results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilizing the deformation or partial superplastic deformation, or both. For adjacent rods in a fuel assembly not to touch at any temperature, the pressure would have to be less than about 1 MPa. These results are compared with those form multirod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behavior of fuel elements in a loss-of-coolant accident are outlined. 37 refs.

  7. Contamination of a PWR primary circuit by fuel pins with failed cladding

    International Nuclear Information System (INIS)

    Janvier, J.C.; Chagrot, M.

    1979-01-01

    The safety authorities in the principal nuclear countries appear to be attaching increasing importance to keeping reactor primary circuits as contamination-free as possible. Therefore, the consequences of cladding failures and especially of those resulting from fabrication defects have to be evaluated, for when these failures become systematic in nature they constitute an important source of contamination in pressurized-water reactors. The Grenoble Nuclear Research Centre is implementing a programme on the study of such failures with a view to analysing the behaviour of failed fuel elements. A distinction is made between two types of cladding failure, depending on whether the primary water enters the fuel pin as soon as the circuits are pressurized (fabrication defect) or whether the failure is caused during operation. The emission of gaseous fission products and halogens has been analysed in different operating modes (steady-state or transient), and in spite of the complexity of the phenomena involved, some results have been obtained which already enable one to evaluate fission product contamination of the primary circuit. (author)

  8. Strategies of management of the nuclear fuel

    International Nuclear Information System (INIS)

    Leon, J.R.; Perez, A.; Filella, J.M.

    1996-01-01

    The management of nuclear fuel is depending on several factors: - Regulatory commission. The enterprises owner of the NPPs.The enterprise owner of the energy distribution. These factors are considered for the management of nuclear fuel. The design of fuel elements, the planning of cycles, the design of core reactors and the costs are analyzed. (Author)

  9. Equipment for nondestructive testing of the PWR and BWR spept fUel elements and assemblies in the NPP storage pools

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1983-01-01

    Design features are considered of units for nondestructive testing of spent fUel elements and fuel assemblies (FA) in the storage pools of NPP with the PWR and BWR reactors. Units for remote viewing control of fuel element cans and FA, for direct measurements of their geometrical dimensions, for FA leak-testing, fuel element can nondestructive testing and gamma scanning, for measuring gaseous fission product pressure and fuel element free volume are described along with units for complex checking of fuel element and FA parameters. The units for nondestructive testing of spent fuel elements and EA are shown to differ both in their designs and a number of checked parameters of fuel elements and FA. The remote viewing and those for measuring the basic FA parameters are most generally employed. Units for complex testing of multiple fuel element parameters, designed in the last few years, are intended for operation with FA disassembled partially or fully and are characteristic of a high degree of computer measuring automation both for the process control and data processing

  10. 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

  11. Power distribution and fuel depletion calculation for a PWR, using LEOPARD and CITATION codes

    International Nuclear Information System (INIS)

    Batista, J.L.

    1982-01-01

    By modifying LEOPARD a new program, LEOCIT, has been developed in which additional subroutines prepare cross-section libraries in 1, 2 or 4 energy groups and subsequently record these on disc or tape in a format appropriate for direct input to the CITATION code. Use of LEOCIT in conjunction with CITATION is demonstrated by simulating the first depletion cycle of Angra Unit 1. In these calculations two energy groups are used in 1/4, X - Y geometry to give the soluble boron curve, the fuel depletion and the point to point power distribution in Angra 1. Finally relevant results obtained here are compared with those published by Westinghouse, CNEN and Furnas and recommendations are made to improve the system of neutronic calculation developed in this work. (Author) [pt

  12. Conservatism in the actinide-only burnup credit for PWR spent nuclear fuel packages

    International Nuclear Information System (INIS)

    Lancaster, D.B.; Rahimi, M.; Thornton, J.

    1996-01-01

    In May 1995, the U.S. Department of Energy (DOE) submitted a topical report to the U.S. Nuclear Regulatory Commission (NRC) to gain actinide-only burnup credit for spent nuclear fuel (SNF) storage, transportation, or disposal packages. After approval of this topical report, DOE intends further submittals to the NRC to acquire additional burnup credit (e.g., the topical does not use fission products and is limited to only the first 100 yr of disposal). The NRC has responded to the topical with its preliminary questions. To aid in evaluation of the method, a review of the conservatism in the actinide-only burnup credit methodology was performed. An overview of the actinide-only burnup credit methodology is presented followed by a summary of the conservatism

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-31

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

  14. Assessment and management of ageing of major nuclear power plant components important to safety: PWR pressure vessels. 2007 update

    International Nuclear Information System (INIS)

    2007-06-01

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that effective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) is one of the most important issues for plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. IAEA-TECDOC-1120 documented ageing assessment and management practices for pressurized water reactor (PWR) reactor pressure vessels (RPVs) that were current at the time of its finalization in 1997-1998. Safety significant operating events have occurred since the finalization of the TECDOC, e.g. primary water stress corrosion cracking (PWSCC) of Alloy 600 control rod drive mechanism (CRDM) penetrations and boric acid corrosion/wastage of RPV heads, which threatened the integrity of the RPV heads. These events led to new ageing management actions by both NPP operators and regulators. Therefore it was recognized that IAEA-TECDOC-1120 should be updated by incorporating those new events and their countermeasures. The objective of this report is to update IAEA-TECDOC-1120 in order to provide current ageing management guidance for PWR RPVs to all involved in the operation and regulation of PWRs and thus to help ensure PWR RPV integrity in IAEA Member States throughout their entire service life

  15. Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals: 2007 update

    International Nuclear Information System (INIS)

    2007-06-01

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that effective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) is one of the most important issues for plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wearout of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. IAEA-TECDOC-1119 documents ageing assessment and management practices for PWR Reactor Vessel Internals (RVIs) that were current at the time of its finalization in 1997-1998. Safety significant operating events have occurred since the finalization of the TECDOC, e.g. irradiation assisted stress corrosion cracking (IASCC) of baffle-former bolts, which threatened the integrity of the vessel internals. In addition, concern of fretting wear of control rod guide tubes has been raised in Japan. These events led to new ageing management actions by both NPP operators and regulators. Therefore it was recognized that IAEA-TECDOC-1119 should be updated by incorporating those new events and their countermeasures. The objective of this report is to update relevant sections of the existing IAEA-TECDOC- 1119 in order to provide current ageing management guidance for PWR RVIs to all involved in the operation and regulation of PWRs and thus to help ensure PWR safety in IAEA Member States throughout their entire service life

  16. Spent Nuclear Fuel Project Safety Management Plan

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1996-02-01

    The Spent Nuclear Fuel Project Safety Management Plan describes the new nuclear facility regulatory requirements basis for the Spemt Nuclear Fuel (SNF) Project and establishes the plan to achieve compliance with this basis at the new SNF Project facilities

  17. Reactivity and isotopic composition of spent PWR [pressurized-water-reactor] fuel as a function of initial enrichment, burnup, and cooling time

    International Nuclear Information System (INIS)

    Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

    1987-10-01

    This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub ∞/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub ∞/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub ∞/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs

  18. Irradiation temperature memorization by retention of krypton-85. Application to the temperature determination for the internal cladding surface of fuel elements in PWR

    International Nuclear Information System (INIS)

    Fremiot, Claude

    1977-01-01

    The temperature of the inner surface of the cladding fuel elements, which can not be measured directly, can be determined after irradiation. During its stage within the reactor, the cladding is bombarded by krypton-85 fission product, which is trapped in the metallic lattice defects. The experience shows that the krypton release during postirradiation heating takes place at the irradiation temperature. This method was applied for PWR fuel element. A very simple model for retention and release of the krypton is proposed. The krypton trap-energy in zircaloy partakes in this model. This technique can be ordered amongst the Hot'Lab' control methods and expert appraisements. It is pointed out that the principal interest in that method is the fact that it does not need any fuel element instrumentation. At the present, this method is being used by CEA for routine-control. [fr

  19. An experimental and analytical study of fluid flow and critical heat flux in PWR fuel elements

    International Nuclear Information System (INIS)

    Bowditch, F.H.; Mogford, D.J.

    1987-02-01

    This report describes experiments that have been carried out at the Winfrith Establishment of the United Kingdom Atomic Energy Authority to determine the critical heat flux characteristics of pressurized water reactor fuel elements over an unusually wide range of coolant flow conditions that are relevant to both normal and fault conditions of reactor operation. The experiments were carried out in the TITAN loop using an electrically heated bundle of 25 rods of 9.5 mm diameter on a 12.7 mm pitch fitted with plain grids in order to provide a generic base for code validation. The fully tabulated experimental data for critical heat flux, pressure drop and sub-channel mixing are encompassed by ranges of pressure between 20 and 160 Bar, coolant flow between 150 and 3600 Kg/m 2 s, and coolant inlet temperature between 150 and 320 0 C. The results of the experiments are compared with predicted data based upon several established critical heat flux correlations. It is concluded that the extrapolation of some correlations to conditions beyond their intended range of application can lead to dangerous over estimates of critical heat flux, but the Winfrith WSC-2 and the EPRI NP-2609 correlations perform well over the whole data range and correlate all data with RMS errors of 9% and 6% respectively. (author)

  20. Monte Carlo simulation of the electron and X-ray depth distribution for quantitative electron probe microanalysis of PWR spent fuels

    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)

    2011-10-15

    Electron probe microanalysis requires several corrections to quantify an element of a specimen. The X-rays produced by the primary beam are created at some depth in the specimen. This distribution is usually represented as the function {Phi}(pz), and it is possible to calculate the correction factors for atomic number and absorption effects. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to quantify some elements of the PWR spent fuel with 50 GWd/tU of burnup and 2 years of cooling time