New Routes to Lanthanide and Actinide Nitrides

Energy Technology Data Exchange (ETDEWEB)

Butt, D.P.; Jaques, B.J.; Osterberg, D.D. [Boise State University, 1910 University Dr., Boise, Idaho 83725-2075 (United States); Marx, B.M. [Concurrent Technologies Corporation, Johnstown, PA (United States); Callahan, P.G. [Carnegie Mellon University, Pittsburgh, PA (United States); Hamdy, A.S. [Central Metallurgical R and D Institute, Helwan, Cairo (Egypt)

2009-06-15

The future of nuclear energy in the U.S. and its expansion worldwide depends greatly on our ability to reduce the levels of high level waste to minimal levels, while maintaining proliferation resistance. Implicit in the so-called advanced fuel cycle is the need for higher levels of fuel burn-up and consequential use of complex nuclear fuels comprised of fissile materials such as Pu, Am, Np, and Cm. Advanced nitride fuels comprised ternary and quaternary mixtures of uranium and these actinides have been considered for applications in advanced power plants, but there remain many processing challenges as well as necessary qualification testing. In this presentation, the advantages and disadvantages of nitride fuels are discussed. Methods of synthesizing the raw materials and sintering of fuels are described including a discussion of novel, low cost routes to nitrides that have the potential for reducing the cost and footprint of a fuel processing plant. Phase pure nitrides were synthesized via four primary methods; reactive milling metal flakes in nitrogen at room temperature, directly nitriding metal flakes in a pure nitrogen atmosphere, hydriding metal flakes prior to nitridation, and carbo-thermically reducing the metal oxide and carbon mixture prior to nitridation. In the present study, the sintering of UN, DyN, and their solid solutions (U{sub x}, Dy{sub 1-x}) (x = 1 to 0.7) were also studied. (authors)

Estimate of fuel burnup spatial a multipurpose reactor in computer simulation; Estimativa da queima espacial do combustivel de um reator multiproposito por simulacao computacional

Energy Technology Data Exchange (ETDEWEB)

Santos, Nadia Rodrigues dos, E-mail: nadia.santos@ifrj.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: malu@ien.gov.br, E-mail: zrlima@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2015-07-01

In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)

Burnup Estimation of Rhodium Self-Powered Neutron Detector Emitter in VVER Reactor Core Using Monte Carlo Simulations

OpenAIRE

Khrutchinsky, А. А.; Kuten, S. A.; Babichev, L. F.

2011-01-01

Estimation of burn-up in a rhodium-103 emitter of self-powered neutron detector in VVER-1000 reactor core has been performed using Monte Carlo simulations within approximation of a constant neutron flux.

Analysis of high burnup fuel safety issues

Energy Technology Data Exchange (ETDEWEB)

Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

2000-12-01

Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

Analysis of high burnup fuel safety issues

International Nuclear Information System (INIS)

Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

2000-12-01

Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development

Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations

International Nuclear Information System (INIS)

Wagner, J.C.; DeHart, M.D.

2000-01-01

This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ''end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified

Microstructural evolution during nitriding, finite element simulation and experimental assessment

Energy Technology Data Exchange (ETDEWEB)

Hassani-Gangaraj, S.M. [Politecnico di Milano, Dipartimento di Meccanica, Via La Masa, 1, 20156 Milano (Italy); Guagliano, M., E-mail: mario.guagliano@polimi.it [Politecnico di Milano, Dipartimento di Meccanica, Via La Masa, 1, 20156 Milano (Italy)

2013-04-15

A finite element simulation of nitriding is proposed in this paper, using the analogy between diffusion and heat conduction, to overcome the shortcomings of the classical internal oxidation model in predicting the kinetics of layer growth and nitrogen distribution during nitriding. To verify the model, a typical gas nitriding has been carried out on an axisymmetric specimen. Treated specimen has been characterized using optical microscopy (OM), scanning electron microscopy (SEM), micro-hardness and X-Ray diffraction (XRD) measurements. It was found that the so-called diffusion zone can be divided into two parts with different influence on the mechanical characteristics including residual stress and hardening. First layer which is a two phase region of ferritic matrix and γ′ (Fe{sub 4}N) makes further improvement with respect to the second layer which is a solid solution of nitrogen in ferrite. The formation of that two phase region, which is not predicted by classical model, can be efficiently recognized by the proposed model. It is also proved that the model has the ability to consider the geometry dependency of layer growth and formation in nitriding.

A microcomputer program for coupled cycle burnup calculations

International Nuclear Information System (INIS)

Driscoll, M.J.; Downar, T.J.; Taylor, E.L.

1986-01-01

A program, designated BRACC (Burnup, Reactivity, And Cycle Coupling), has been developed for fuel management scoping calculations, and coded in the BASIC language in an interactive format for use with microcomputers. BRACC estimates batch and cycle burnups for sequential reloads for a variety of initial core conditions, and permits the user to specify either reload batch properties (enrichment, burnable poison reactivity) or the target cycle burnup. Most important fuel management tactics (out-in or low-leakage loading, coastdown, variation in number of assemblies charged) can be simulated

Physical models for high burnup fuel

International Nuclear Information System (INIS)

Kanyukova, V.; Khoruzhii, O.; Likhanskii, V.; Solodovnikov, G.; Sorokin, A.

2003-01-01

In this paper some models of processes in high burnup fuel developed in Src of Russia Troitsk Institute for Innovation and Fusion Research are presented. The emphasis is on the description of the degradation of the fuel heat conductivity, radial profiles of the burnup and the plutonium accumulation, restructuring of the pellet rim, mechanical pellet-cladding interaction. The results demonstrate the possibility of rather accurate description of the behaviour of the fuel of high burnup on the base of simplified models in frame of the fuel performance code if the models are physically ground. The development of such models requires the performance of the detailed physical analysis to serve as a test for a correct choice of allowable simplifications. This approach was applied in the SRC of Russia TRINITI to develop a set of models for the WWER fuel resulting in high reliability of predictions in simulation of the high burnup fuel

Burnup measurement study and prototype development in HTR-PM

International Nuclear Information System (INIS)

Yan Weihua; Zhang Zhao; Xiao Zhigang; Zhang Liguo

2014-01-01

In a pebble-bed core which employs the multi-pass scheme, it is mandatory to determine the burnup of each pebble after the pebble has been extracted from the core in order to determine whether its design burnup has been reached or whether it has to be reinserted into the core again. The burnup of the fuel pebbles can be determined by measuring the activity of 137 Cs with an HPGe detector because of their good correspondence, which is independent of the irradiation history in the core. Based on experiments and Geant4 simulation, the correction factor between the fuel and calibration source was derived by using the efficiency transfer method. By optimizing spectrum analysis algorithm and parameters, the relative standard deviation of the 137 Cs activity can be still controlled below 3.0% despite of the presence of interfering peaks. On the foundation of the simulation and experiment research, a complete solution for burnup measurement system in HTR-PM is provided. (authors)

Influence of FIMA burnup on actinides concentrations in PWR reactors

Directory of Open Access Journals (Sweden)

Oettingen Mikołaj

2016-01-01

Full Text Available In the paper we present the study on the dependence of actinides concentrations in the spent nuclear fuel on FIMA burnup. The concentrations of uranium, plutonium, americium and curium isotopes obtained in numerical simulation are compared with the result of the post irradiation assay of two spent fuel samples. The samples were cut from the fuel rod irradiated during two reactor cycles in the Japanese Ohi-2 Pressurized Water Reactor. The performed comparative analysis assesses the reliability of the developed numerical set-up, especially in terms of the system normalization to the measured FIMA burnup. The numerical simulations were preformed using the burnup and radiation transport mode of the Monte Carlo Continuous Energy Burnup Code – MCB, developed at the Department of Nuclear Energy, Faculty of Energy and Fuels of AGH University of Science and Technology.

Development of methods for burn-up calculations for LWR's

International Nuclear Information System (INIS)

Jaschik, W.

1978-01-01

This method is based on all burn-up depending data, namely particle densities and neutron spectra, being available in a burn-up library. This one is created by means of a small number of cell burn-up calculations which can easily be carried out and in which the heterogeneous cell structure and self-shielding effects can explicitly be accounted for. Then the cluster burn-up is simulated by adequate correlation of the burn-up data. The advantage of this method is given by - an exact determination of the real spectrum distribution in the individual fuel element clusters; - an exact determination of the burn-up related spectrum variations for each fuel rod and for each burn-up value obtained; - accounting for heterogeneity of the fuel rod cells and the self-shielding in the fuel; high accuracy of the results of a comparably low effort and - simple handling by largely automating the process of computation. Programed realization was achieved by establishing the RSYST modules ABRAJA, MITHOM, and SIMABB and their implementation within the code system. (orig./HP) [de

Kr-85m activity as burnup measurement indicator in a pebble bed reactor based on ORIGEN2.1 Computer Simulation

Science.gov (United States)

Husnayani, I.; Udiyani, P. M.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

Pebble Bed Reactor (PBR) is a high temperature gas-cooled reactor which employs graphite as a moderator and helium as a coolant. In a multi-pass PBR, burnup of the fuel pebble must be measured in each cycle by online measurement in order to determine whether the fuel pebble should be reloaded into the core for another cycle or moved out of the core into spent fuel storage. One of the well-known methods for measuring burnup is based on the activity of radionuclide decay inside the fuel pebble. In this work, the activity and gamma emission of Kr-85m were studied in order to investigate the feasibility of Kr-85m as burnup measurement indicator in a PBR. The activity and gamma emission of Kr-85 were estimated using ORIGEN2.1 computer code. The parameters of HTR-10 were taken as a case study in performing ORIGEN2.1 simulation. The results show that the activity revolution of Kr-85m has a good relationship with the burnup of the pebble fuel in each cycle. The Kr-85m activity reduction in each burnup step,in the range of 12% to 4%, is considered sufficient to show the burnup level in each cycle. The gamma emission of Kr-85m is also sufficiently high which is in the order of 1010 photon/second. From these results, it can be concluded that Kr-85m is suitable to be used as burnup measurement indicator in a pebble bed reactor.

Time step length versus efficiency of Monte Carlo burnup calculations

International Nuclear Information System (INIS)

Dufek, Jan; Valtavirta, Ville

2014-01-01

Highlights: • Time step length largely affects efficiency of MC burnup calculations. • Efficiency of MC burnup calculations improves with decreasing time step length. • Results were obtained from SIE-based Monte Carlo burnup calculations. - Abstract: We demonstrate that efficiency of Monte Carlo burnup calculations can be largely affected by the selected time step length. This study employs the stochastic implicit Euler based coupling scheme for Monte Carlo burnup calculations that performs a number of inner iteration steps within each time step. In a series of calculations, we vary the time step length and the number of inner iteration steps; the results suggest that Monte Carlo burnup calculations get more efficient as the time step length is reduced. More time steps must be simulated as they get shorter; however, this is more than compensated by the decrease in computing cost per time step needed for achieving a certain accuracy

Burn-up measurement in the HTR-module-reactor

International Nuclear Information System (INIS)

Gerhards, E.

1993-05-01

The burn-up status of spherical HTR-fuel elements is determined by a γ-spectrometric analysis of Cs-137 activity. The γ-spectrum recorded by a semiconductor detector up to now is analyzed by complex mathematical and time-consuming methods. For the operation of the HTR-Module-Reactor, however, a fast evaluation of the burn-up status is necessary. It is shown that this can be ensured by a comparison between the measured spectra and simulation results. Using the computer-program HTROGEN and the program system SPECCALC especially developed for this problem the γ-spectra are evaluated as a function of the burn-up status. The method is applied to results available from the operation of the AVR-reactor. The burn-up status determined with different methods corresponds very well within the limits of accuracy. (orig.)

Calculation of isotope burn-up and change in efficiency of absorbing elements of WWER-1000 control and protection system during burn-up

International Nuclear Information System (INIS)

Timofeeva, O.A.; Kurakin, K.U.

2006-01-01

The report deals with fast and thermal neutron flows distribution in structural elements of WWER-1000 fuel assembly and absorbing rods, determination of absorbing isotope burn-up and worth variation in WWER reactor control and protection system rods. Simulation of absorber rod burn-up is provided using code package SAPPHIRE 9 5 end RC W WER allowing detailed description of the core segment spatial model. Maximum burn-up of absorbing rods and respective worth variation of control and protection system rods is determined on the basis of a number of calculations considering known characteristics of fuel cycles (Authors)

Parallel GPU implementation of PWR reactor burnup

International Nuclear Information System (INIS)

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

2016-01-01

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

Choosing the optimum burnup

International Nuclear Information System (INIS)

Geller, L.; Goldstein, L.; Franks, W.A.

1986-01-01

This paper reviews some of the considerations utilities must evaluate when going to higher discharge burnups. The advantages and disadvantages of higher discharge burnups are described, as well as a consistent approach for evaluating optimum discharge burnup and its comparison to current practice. When an analysis is performed over the life of the plant, the design of the terminal cycles has significant impact on the lifetime savings from higher burnups. Designs for high burnup cycles have a greater average inventory value in the core. As one goes to higher burnup, there is a greater likelihood of discarding a larger value in unused fuel unless the terminal cycles are designed carefully. This effect can be large enough in some cases to wipe out the lifetime cost savings relative to operating with a higher discharge burnup cycle

Simulated LOCA Test and Characterization Study Related to High Burn-Up Issue

International Nuclear Information System (INIS)

Park, D. J.; Jung, Y. I.; Choi, B. K.; Park, S. Y.; Kim, H. G.; Park, J. Y.

2012-01-01

For the safety evaluation of fuel cladding during the injection of emergency core coolant, simulated Loss-of-coolant accident (LOCA) test was performed by using Zircaloy-4 fuel cladding samples. Zircaloy-4 tube samples with and without prehydring were oxidized in a steam environment with the test temperature of 1200 .deg. C. Prehydrided cladding was prepared from as-fabricated Zircaloy-4 to study the effects of hydrogen on mechanical properties of cladding during high temperature oxidation and quench conditions. In order to measure the ductility of the tube samples embrittled by quenching water, ring compression test was carried out by using 8 mm ring sample sectioned from oxidized tube sample and microstructural analysis was also performed after simulated LOCA test. The results showed that hydrogen increases oxygen solubility and pickup rate in the beta layer. This reduces ductility of prehydrided fuel cladding compared with as-fabricated cladding. Trend in ductility decrease for prehydrided sample under simulated LOCA condition was very similar with data obtained from tests conducted using irradiated high burn-up fuel claddings

Modeling the Gas Nitriding Process of Low Alloy Steels

Science.gov (United States)

Yang, M.; Zimmerman, C.; Donahue, D.; Sisson, R. D.

2013-07-01

The effort to simulate the nitriding process has been ongoing for the last 20 years. Most of the work has been done to simulate the nitriding process of pure iron. In the present work a series of experiments have been done to understand the effects of the nitriding process parameters such as the nitriding potential, temperature, and time as well as surface condition on the gas nitriding process for the steels. The compound layer growth model has been developed to simulate the nitriding process of AISI 4140 steel. In this paper the fundamentals of the model are presented and discussed including the kinetics of compound layer growth and the determination of the nitrogen diffusivity in the diffusion zone. The excellent agreements have been achieved for both as-washed and pre-oxided nitrided AISI 4140 between the experimental data and simulation results. The nitrogen diffusivity in the diffusion zone is determined to be constant and only depends on the nitriding temperature, which is ~5 × 10-9 cm2/s at 548 °C. It proves the concept of utilizing the compound layer growth model in other steels. The nitriding process of various steels can thus be modeled and predicted in the future.

Whole core burnup calculations using `MCNP`

Energy Technology Data Exchange (ETDEWEB)

Haran, O; Shaham, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

1996-12-01

Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors).

Whole core burnup calculations using 'MCNP'

International Nuclear Information System (INIS)

Haran, O.; Shaham, Y.

1996-01-01

Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors)

New burnup calculation of TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.

2015-01-01

The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

Lattice cell burnup calculation

International Nuclear Information System (INIS)

Pop-Jordanov, J.

1977-01-01

Accurate burnup prediction is a key item for design and operation of a power reactor. It should supply information on isotopic changes at each point in the reactor core and the consequences of these changes on the reactivity, power distribution, kinetic characters, control rod patterns, fuel cycles and operating strategy. A basic stage in the burnup prediction is the lattice cell burnup calculation. This series of lectures attempts to give a review of the general principles and calculational methods developed and applied in this area of burnup physics

BEAVRS full core burnup calculation in hot full power condition by RMC code

International Nuclear Information System (INIS)

Liu, Shichang; Liang, Jingang; Wu, Qu; Guo, JuanJuan; Huang, Shanfang; Tang, Xiao; Li, Zeguang; Wang, Kan

2017-01-01

Highlights: • TMS and thermal scattering interpolation were developed to treat cross sections OTF. • Hybrid coupling system was developed for HFP burnup calculation of BEAVRS benchmark. • Domain decomposition was applied to handle memory problem of full core burnup. • Critical boron concentration with burnup by RMC agrees with the benchmark results. • RMC is capable of multi-physics coupling for simulations of nuclear reactors in HFP. - Abstract: Monte Carlo method can provide high fidelity neutronics analysis of different types of nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with multi-physics interacting and coupling. MC codes can couple with depletion solver and thermal-hydraulics (T/H) codes simultaneously for the “transport-burnup-thermal-hydraulics” coupling calculations. MIT BEAVRS is a typical “transport-burnup-thermal-hydraulics” coupling benchmark. In this paper, RMC was coupled with sub-channel code COBRA, equipped with on-the-fly temperature-dependent cross section treatment and large-scale detailed burnup calculation based on domain decomposition. Then RMC was applied to the full core burnup calculations of BEAVRS benchmark in hot full power (HFP) condition. The numerical tests show that domain decomposition method can achieve the consistent results compared with original version of RMC while enlarging the computational burnup regions. The results of HFP by RMC agree well with the reference values of BEAVRS benchmark and also agree well with those of MC21. This work proves the feasibility and accuracy of RMC in multi-physics coupling and lifecycle simulations of nuclear reactors.

A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

Energy Technology Data Exchange (ETDEWEB)

Makmal, T. [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel); Nuclear Physics and Engineering Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Aviv, O. [Radiation Safety Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Gilad, E., E-mail: gilade@bgu.ac.il [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel)

2016-10-21

A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections. - Highlights: • Simple, inexpensive, safe and flexible experimental setup that can be quickly deployed. • Experimental results are thoroughly corroborated against ORIGEN2 burnup code. • Experimental uncertainty of 9% and 5% deviation between measurements and simulations. • Very high burnup MTR fuel element is examined, with 60% depletion of {sup 235}U. • Impact of highly irregular irradiation regime on burnup evaluation is studied.

New approach to derive linear power/burnup history input for CANDU fuel codes

International Nuclear Information System (INIS)

Lac Tang, T.; Richards, M.; Parent, G.

2003-01-01

The fuel element linear power / burnup history is a required input for the ELESTRES code in order to simulate CANDU fuel behavior during normal operating conditions and also to provide input for the accident analysis codes ELOCA and SOURCE. The purpose of this paper is to present a new approach to derive 'true', or at least more realistic linear power / burnup histories. Such an approach can be used to recreate any typical bundle power history if only a single pair of instantaneous values of bundle power and burnup, together with the position in the channel, are known. The histories obtained could be useful to perform more realistic simulations for safety analyses for cases where the reference (overpower) history is not appropriate. (author)

Comparative study on plutonium and MA recycling in equilibrium burnup and standard burnup of PWR

International Nuclear Information System (INIS)

Waris, Abdul; Kurniadi, Rizal; Su'ud, Zaki; Permana, Sidik

2005-01-01

The equilibrium burnup model is a powerful method since its can handle all possible generated nuclides in any nuclear system. Moreover, this method is a simple time independent method. Hence the equilibrium burnup method could be very useful for evaluating and forecasting the characteristics of any nuclear fuel cycle, even the strange one, e.g. all nuclides are confined in the reactor. However, this method needs to be verified since the method is not a standard tool. The present study aimed to compare the characteristics of plutonium recycling and plutonium and minor actinides (MA) recycling in PWR with the equilibrium burnup and the standard burnup. In order to become more comprehensive study, an influence of moderator-to-fuel volume ratio (MFR) changes by changing the pin-pitch of fuel cell has been evaluated. The MFR ranges from 0.5 to 4.0. For the equilibrium burnup we used equilibrium cell-burnup code. We have employed 1368 nuclides in the equilibrium calculation with 129 of them are heavy metals (HMs). For standard burnup, SRAC2002 code has been utilized with 26 HMs and 66 fission products (FPs). The JENDL 3.2 library has been employed for both burnup schemes. The uranium, plutonium and MA vector, which resulted from the equilibrium burnup are directly used as fuel input composition for the standard burnup calculation. Both burnup results demonstrate that plutonium recycling and plutonium and MA recycling can be conducted safer in tight lattice core. They are also show the similar trend in neutron spectrum, which become harder with the increasing number of recycled heavy nuclides as well as the decreasing of the MFR values. However, there are some discrepancy on the effective multiplication factor and the conversion ratio, especially for the reactor core for MFR ≥ 2.0. (author)

Improvements for Monte Carlo burnup calculation

Energy Technology Data Exchange (ETDEWEB)

Shenglong, Q.; Dong, Y.; Danrong, S.; Wei, L., E-mail: qiangshenglong@tsinghua.org.cn, E-mail: d.yao@npic.ac.cn, E-mail: songdr@npic.ac.cn, E-mail: luwei@npic.ac.cn [Nuclear Power Inst. of China, Cheng Du, Si Chuan (China)

2015-07-01

Monte Carlo burnup calculation is development trend of reactor physics, there would be a lot of work to be done for engineering applications. Based on Monte Carlo burnup code MOI, non-fuel burnup calculation methods and critical search suggestions will be mentioned in this paper. For non-fuel burnup, mixed burnup mode will improve the accuracy of burnup calculation and efficiency. For critical search of control rod position, a new method called ABN based on ABA which used by MC21 will be proposed for the first time in this paper. (author)

Fabrication and testing of uranium nitride fuel for space power reactors

Science.gov (United States)

Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.

1988-02-01

Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

High Burnup Effects Program

International Nuclear Information System (INIS)

Barner, J.O.; Cunningham, M.E.; Freshley, M.D.; Lanning, D.D.

1990-04-01

This is the final report of the High Burnup Effects Program (HBEP). It has been prepared to present a summary, with conclusions, of the HBEP. The HBEP was an international, group-sponsored research program managed by Battelle, Pacific Northwest Laboratories (BNW). The principal objective of the HBEP was to obtain well-characterized data related to fission gas release (FGR) for light water reactor (LWR) fuel irradiated to high burnup levels. The HBEP was organized into three tasks as follows: Task 1 -- high burnup effects evaluations; Task 2 -- fission gas sampling; and Task 3 -- parameter effects study. During the course of the HBEP, a program that extended over 10 years, 82 fuel rods from a variety of sources were characterized, irradiated, and then examined in detail after irradiation. The study of fission gas release at high burnup levels was the principal objective of the program and it may be concluded that no significant enhancement of fission gas release at high burnup levels was observed for the examined rods. The rim effect, an as yet unquantified contributor to athermal fission gas release, was concluded to be the one truly high-burnup effect. Though burnup enhancement of fission gas release was observed to be low, a full understanding of the rim region and rim effect has not yet emerged and this may be a potential area of further research. 25 refs., 23 figs., 4 tabs

Comparison of scale/triton and helios burnup calculations for high burnup LWR fuel

Energy Technology Data Exchange (ETDEWEB)

Tittelbach, S.; Mispagel, T.; Phlippen, P.W. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

2009-07-01

The presented analyses provide information about the suitability of the lattice burnup code HELIOS and the recently developed code SCALE/TRITON for the prediction of isotopic compositions of high burnup LWR fuel. The accurate prediction of the isotopic inventory of high burnt spent fuel is a prerequisite for safety analyses in and outside of the reactor core, safe loading of spent fuel into storage casks, design of next generation spent fuel casks and for any consideration of burnup credit. Depletion analyses are performed with both burnup codes for PWR and BWR fuel samples which were irradiated far beyond 50 GWd/t within the LWR-PROTEUS Phase II project. (orig.)

High burnup issues and modelling strategies

International Nuclear Information System (INIS)

Dutta, B.K.

2005-01-01

The performance of high burnup fuel is affected by a number of phenomena, such as, conductivity degradation, modified radial flux profile, fission gas release from high burnup structures, PCMI, burnup dependent thermo-mechanical properties, etc. The modelling strategies of some of these phenomena are available in literature. These can be readily incorporated in a fuel modelling performance code. The computer code FAIR has been developed in BARC over the years to evaluate the fuel performance at extended burnup and modelling of the fuel rods for advanced fuel cycles. The present paper deals with the high burnup issues in the fuel pins, their modelling strategies and results of the case studies specifically involving high burnup fuel. (author)

Burnup verification tests with the FORK measurement system-implementation for burnup credit

International Nuclear Information System (INIS)

Ewing, R.I.

1994-01-01

Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. It was designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program and is well suited to verify burnup and cooling time records at commercial Pressurized Water Reactor (PWR) sites. This report deals with the application of the FORK system to burnup credit operations

Mass spectrometric study of vaporization of (U,Pu)O2 fuel simulating high burnup

International Nuclear Information System (INIS)

Maeda, Atsushi; Ohmichi, Toshihiko; Fukushima, Susumu; Handa, Muneo

1985-08-01

The vaporization behavior of (U,Pu)O 2 fuel simulatig high burnup was studied in the temperature range of 1,573 -- 2,173 K by high temperature mass spectrometry. The phases in the simulated fuel were examined by X-ray microprobe analysis. The relationship between chemical form and vaporization behavior of simulated fission product elements was discussed. Pd, Sr, Ba, Ce and actinide-bearing vapor species were observed, and it was clarified that Pd vapor originated from metallic inclusion and Sr and Ce vapors, from mixed oxide fuel matrix. The vaporization behavior of the actinide elements was somewhat similar to that of hypostoichiometric mixed oxide fuel. The behavior of Ba-bearing vapor species changed markedly over about 2,000 K. From the determination of BaO vapor pressures over simulated fuel and BaZrO 3 , it was revealed thermodynamically that the transformation of the chemical form of Ba about 2,000 K, i.e., dissolution of BaZrO 3 phase into fuel matrix, might be the reason of the observed vapor pressure change. (author)

Fission gas release and fuel rod chemistry related to extended burnup

International Nuclear Information System (INIS)

1993-04-01

The purpose of the meeting was to review the state of the art in fission gas release and fuel rod chemistry related to extended burnup. The meeting was held in a time when national and international programmes on water reactor fuel irradiated in experimental reactors were still ongoing or had reached their conclusion, and when lead test assemblies had reached high burnup in power reactors and been examined. At the same time, several out-of-pile experiments on high burnup fuel or with simulated fuel were being carried out. As a result, significant progress has been registered since the last meeting, particularly in the evaluation of fuel temperature, the degradation of the global thermal conductivity with burnup and in the understanding of the impact on fission gas release. Fifty five participants from 16 countries and one international organization attended the meeting. 28 papers were presented. A separate abstract was prepared for each of the papers. Refs, figs, tabs and photos

A simulation of the temperature overshoot observed at high burnup in annular fuel pellets

International Nuclear Information System (INIS)

Baron, D.; Couty, J.C.

1997-01-01

Instrumented experiments have been carried out in recent years to calibrate and improve temperature calculations at high burnup in PWR nuclear fuel rods. The introduction of a thermocouple in the fuel stack allows the experiment to record the centre-line temperature all along the irradiation or re-irradiation. The results obtained on fresh fuel have not revealed any abnormal behavior as have observations done on high burnup rods. In this case, a sudden overshoot has been recorded on the thermocouple temperature above an average power threshold. Several hypotheses have been suggested. Only two seem to be acceptable: one in relation to an effect of grain decohesion, another based on a modification of fuel chemistry. The apparent reversibility of the phenomena when power decreases led us to prefer the first explanation. Indeed, the introduction of a thermocouple means that annular fuel pellets must be used. These are either initially manufactured with a central hole or drilled after base irradiation, using the ''RISOE'' technique. One must bear in mind that the use of such annular pellets drastically changes the crack pattern as irradiation proceeds. This is due to a different stress field which, combined with a weakening of the grain binding energy, leads to a partial grain decohesion on the inner face of the annular pellet. Modification of the grain binding energy is related to the presence of an increasing local population of gas bubbles and metallic precipitates at grain boundaries, as swelling creates intergranular local stresses which also could probably enhance the grain decohesion process. This grain decohesion concerns a 250 to 350 μm depth and shows a narrow cracks network through which released fission gas can flow, temporarily pushing the resident helium gas out. The low conductivity of these gaseous fission products and the numerous gas layers created this way could partly explain the unexpected temperatures measured in high burnup fuels. The purpose of

A simplified burnup calculation strategy with refueling in static molten salt reactor

International Nuclear Information System (INIS)

Srivastava, A.K.; Gupta, Anurag; Krishnani, P.D.

2015-01-01

Molten Salt Reactors, by nature can be refuelled and reprocessed online. Thus, a simulation methodology has to be developed which can consider online refueling and reprocessing aspect of the reactor. To cater such needs a simplified burnup calculation strategy to account for refueling and removal of molten salt fuel at any desired burnup has been identified in static molten salt reactor in batch mode as a first step of way forward. The features of in-house code ITRAN has been explored for such calculations. The code also enables us to estimate the reactivity introduced in the system due to removal of any number of considered nuclides at any burnup. The effect of refueling fresh fuel and removal of burned fuel has been studied in batch mode with in-house code ITRAN. The effect of refueling and burnup on change in reactivity per day has been analyzed. The analysis of removal of 233 Pa at a particular burnup has been carried out. The similar analysis has been performed for some other nuclides also. (author)

Value of burnup credit beyond actinides

International Nuclear Information System (INIS)

Lancaster, D.; Fuentes, E.; Kang, Chi.

1997-01-01

DOE has submitted a topical report to the NRC justifying burnup credit based only on actinide isotopes (U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241). When this topical report is approved, it will allow a great deal of the commercial spent nuclear fuel to be transported in significantly higher capacity casks. A cost savings estimate for shipping fuel in 32 assembly (burnup credit) casks as opposed to 24 assembly (non-burnup credit) casks was previously presented. Since that time, more detailed calculations have been performed using the methodology presented in the Actinide-Only Burnup Credit Topical Report. Loading curves for derated casks have been generated using actinide-only burnup credit and are presented in this paper. The estimates of cost savings due to burnup credit for shipping fuel utilizing 32, 30, 28, and 24 assembly casks where only the 24 assembly cask does not burnup credit have been created and are discussed. 4 refs., 2 figs

Burnup-dependent core neutronics analysis of plate-type research reactor using deterministic and stochastic methods

International Nuclear Information System (INIS)

Liu, Shichang; Wang, Guanbo; Liang, Jingang; Wu, Gaochen; Wang, Kan

2015-01-01

Highlights: • DRAGON & DONJON were applied in burnup calculations of plate-type research reactors. • Continuous-energy Monte Carlo burnup calculations by RMC were chosen as references. • Comparisons of keff, isotopic densities and power distribution were performed. • Reasons leading to discrepancies between two different approaches were analyzed. • DRAGON & DONJON is capable of burnup calculations with appropriate treatments. - Abstract: The burnup-dependent core neutronics analysis of the plate-type research reactors such as JRR-3M poses a challenge for traditional neutronics calculational tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity, large leakage and the particular neutron spectrum of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the burnup-dependent core neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON & DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic one. In the first stage, the homogenizations of few-group cross sections by DRAGON and the full core diffusion calculations by DONJON have been verified by comparing with the detailed Monte Carlo simulations. In the second stage, the burnup-dependent calculations of both assembly level and the full core level were carried out, to examine the capability of the deterministic code system DRAGON & DONJON to reliably simulate the burnup-dependent behavior of research reactors. The results indicate that both RMC and DRAGON & DONJON code system are capable of burnup-dependent neutronics analysis of research reactors, provided that appropriate treatments are applied in both assembly and core levels for the deterministic codes

Local heating with titanium nitride nanoparticles

DEFF Research Database (Denmark)

Guler, Urcan; Ndukaife, Justus C.; Naik, Gururaj V.

2013-01-01

We investigate the feasibility of titanium nitride (TiN) nanoparticles as local heat sources in the near infrared region, focusing on biological window. Experiments and simulations provide promising results for TiN, which is known to be bio-compatible.......We investigate the feasibility of titanium nitride (TiN) nanoparticles as local heat sources in the near infrared region, focusing on biological window. Experiments and simulations provide promising results for TiN, which is known to be bio-compatible....

Burnup credit for storage and transportation casks

International Nuclear Information System (INIS)

Wells, A.H.

1988-01-01

The application of burnup credit to storage and transportation cask licensing results in a significant improvement in cask capacity and an associated reduction of the cost per kilogram of uranium in the cask contents. The issues for licensing with burnup credit deal primarily with the treatment of fission product poisons and methods of verification of burnup during cask operations. Other issues include benchmarking of cross-section sets and codes and the effect of spatial variation of burnup within an assembly. The licensing of burnup credit for casks will be complex, although the criticality calculations are not themselves difficult. Attention should be directed to the use of fission product poisons and the uncertainties that they introduce. Verification of burnup by measurements will remove some of the concerns for criticality safety. Calculations for burnup credit casks should consider rod-to-rod and axial variations of burnup, as well as variability of burnable poisons it they are used in the assembly. In spite of the complexity of cask burnup credit licensing issues, these issues appear to be resolvable within the current state of the art of criticality safety

Application of burnup credit concept to transport

International Nuclear Information System (INIS)

Futamura, Yoshiaki; Nakagome, Yoshihiro.

1994-01-01

For the design and safety assessment of the casks for transporting spent fuel, the fuel contained in them has been assumed to be new fuel. The reason is, it was difficult to evaluate the variation of the reactivity of fuel, and the research on the affecting factors and the method of measuring burnup were not much advanced. Recently, high burnup fuel has been adopted, and initial degree of enrichment rose. The research has been advanced for pursuing the economy of the casks for spent fuel, and burnup credit has become applicable to their design and safety assessment. As the result, the containing capacity increases by about 20%. When burnup credit is considered, it is necessary to confirm accurately the burnup of spent fuel. The burnup dependence of the concentration of fissile substances and neutron emissivity, the coolant void dependence of the concentration of fissile substances, and the relation of neutron multiplication rate with initial degree of enrichment or burnup are discussed. The conceptual design of casks considering burnup credit and its assessment, the merit, problem and the countermeasures to it when burnup credit is introduced are described. (K.I.)

Modeling CANDU type fuel behaviour during extended burnup irradiations using a revised version of the ELESIM code

International Nuclear Information System (INIS)

Arimescu, V.I.; Richmond, W.R.

1992-05-01

The high-burnup database for CANDU fuel, with a variety of cases, offers a good opportunity to check models of fuel behaviour, and to identify areas for improvement. Good agreement of calculated values of fission-gas release, and sheath hoop strain, with experimental data indicates that the global behaviour of the fuel element is adequately simulated by a computer code. Using, the ELESIM computer code, the fission-gas release, swelling, and fuel pellet expansion models were analysed, and changes made for gaseous swelling, and diffusional release of fission-gas atoms to the grain boundaries. Using this revised version of ELESIM, satisfactory agreement between measured values of fission-gas release was found for most of the high-burnup database cases. It is concluded that the revised version of the ELESIM code is able to simulate with reasonable accuracy high-burnup as well as low-burnup CANDU fuel

Threshold burnup for recrystallization and model for rim porosity in the high burnup UO2 fuel

International Nuclear Information System (INIS)

Lee, Byung Ho; Koo, Yang Hyun; Sohn, Dong Seong

1998-01-01

Applicability of the threshold burnup for rim formation was investigated as a function of temperature by Rest's model. The threshold burnup was the lowest in the intermediate temperature region, while on the other temperature regions the threshold burnup is higher. The rim porosity was predicted by the van der Waals equation based of the rim pore radius of 0.75μm and the overpressurization model on rim pores. The calculated centerline temperature is in good agreement with the measured temperature. However, more efforts seem to be necessary for the mechanistic model of the rim effect including rim growth with the fuel burnup

Alloy Effects on the Gas Nitriding Process

Science.gov (United States)

Yang, M.; Sisson, R. D.

2014-12-01

Alloy elements, such as Al, Cr, V, and Mo, have been used to improve the nitriding performance of steels. In the present work, plain carbon steel AISI 1045 and alloy steel AISI 4140 were selected to compare the nitriding effects of the alloying elements in AISI 4140. Fundamental analysis is carried out by using the "Lehrer-like" diagrams (alloy specific Lehrer diagram and nitriding potential versus nitrogen concentration diagram) and the compound layer growth model to simulate the gas nitriding process. With this method, the fundamental understanding for the alloy effect based on the thermodynamics and kinetics becomes possible. This new method paves the way for the development of new alloy for nitriding.

Burnup determination of mass spectrometry for nuclear fuels

International Nuclear Information System (INIS)

Zhang Chunhua.

1987-01-01

The various methods currently being used in burnup determination of nuclear fuels are studied and reviewed. The mass spectrometry method of destructive testing is discussed emphatically. The burnup determination of mass spectrometry includes heavy isotopic abundance ratio method and isotope dilution mass spectrometry used as burnup indicator for the fission products. The former is applied to high burnup level, but the later to various burnup level. According to experiences, some problems which should be noticed in burnup determination of mass spectrometry are presented

Study of the neutronic performances of cores with mixed nitride fuel [(U,Pu)N] for fast neutron reactors

International Nuclear Information System (INIS)

Merzouk, Hamid

1992-01-01

This paper proposes a core design of fast reactor using mixed nitride fuel [(U,Pu)N], having small loss of reactivity and reaching a maximum thermal burn-up rate from 150 GWd/t, while being managed in single batch (renewal of the fuel in only one time for all the subassemblies of the core). This work was completed with aid of the studies of sensibilities of the fast reactors cores to principal parameters: general design of the core, volumetric percentages of the various mixture of materials composing the core, initial enrichments of the fuel. A detailed optimization study on the selected core was conducted complying with safety criteria taking into consideration of consequences of nitride material presence on fuel assembly design rules. (author) [fr

On-line extraction of the variance caused by burn-up in in-core three-dimensional power distribution

International Nuclear Information System (INIS)

Wang Yaqi; Luo Zhengpei; Li Fu; Liu Wenfeng

2001-01-01

In most of PWRs, the ex-core ion-chambers are the sole real-time sensors to respond to in-core power and its axial offset. However, the calibration coefficient of the ion-chambers depends on the (3D) power distribution and varies with the burn-up. People expect to know the variance in distribution caused by burn-up directly from the signals of ion-chambers. This expectation is not realized as yet, because an ion-chamber almost only responds to its nearest fuel assemblies. The authors then developed a two-step method for burn-up characteristic extraction: the harmonics synthesis method and harmonics' burn-up grouping. Using the extracted burn-up characteristics, the relationship between the readings of the ex-core ion-chambers and the in-core 3D power distribution is set up. Through the simulation on the heating reactor, the method of burn-up characteristic extraction is verified under engineering conditions. It is possible to on-line extract the variance caused by burn-up in 3D power distribution

Encapsulation of cisplatin as an anti-cancer drug into boron-nitride and carbon nanotubes: Molecular simulation and free energy calculation

Energy Technology Data Exchange (ETDEWEB)

Roosta, Sara [Molecular Simulation Research Laboratory, Department of Chemistry, Iran University of Science & Technology, Tehran (Iran, Islamic Republic of); Hashemianzadeh, Seyed Majid, E-mail: hashemianzadeh@iust.ac.ir [Molecular Simulation Research Laboratory, Department of Chemistry, Iran University of Science & Technology, Tehran (Iran, Islamic Republic of); Ketabi, Sepideh, E-mail: sepidehketabi@yahoo.com [Department of Chemistry, East Tehran Branch, Islamic Azad University, Tehran (Iran, Islamic Republic of)

2016-10-01

Encapsulation of cisplatin anticancer drug into the single walled (10, 0) carbon nanotube and (10, 0) boron-nitride nanotube was investigated by quantum mechanical calculations and Monte Carlo Simulation in aqueous solution. Solvation free energies and complexation free energies of the cisplatin@ carbon nanotube and cisplatin@ boron-nitride nanotube complexes was determined as well as radial distribution functions of entitled compounds. Solvation free energies of cisplatin@ carbon nanotube and cisplatin@ boron-nitride nanotube were − 4.128 kcal mol{sup −1} and − 2457.124 kcal mol{sup −1} respectively. The results showed that cisplatin@ boron-nitride nanotube was more soluble species in water. In addition electrostatic contribution of the interaction of boron- nitride nanotube complex and solvent was − 281.937 kcal mol{sup −1} which really more than Van der Waals and so the electrostatic interactions play a distinctive role in the solvation free energies of boron- nitride nanotube compounds. On the other hand electrostatic part of the interaction of carbon nanotube complex and solvent were almost the same as Van der Waals contribution. Complexation free energies were also computed to study the stability of related structures and the free energies were negative (− 374.082 and − 245.766 kcal mol{sup −1}) which confirmed encapsulation of drug into abovementioned nanotubes. However, boron-nitride nanotubes were more appropriate for encapsulation due to their larger solubility in aqueous solution. - Highlights: • Solubility of cisplatin@ boron-nitride nanotube is larger than cisplatin@ carbon nanotube. • Boron- nitride nanotube complexes have larger electrostatic contribution in solvation free energy. • Complexation free energies confirm encapsulation of drug into the nanotubes in aqueous solution. • Boron- nitride nanotubes are appropriate drug delivery systems compared with carbon nanotubes.

Encapsulation of cisplatin as an anti-cancer drug into boron-nitride and carbon nanotubes: Molecular simulation and free energy calculation

International Nuclear Information System (INIS)

Roosta, Sara; Hashemianzadeh, Seyed Majid; Ketabi, Sepideh

2016-01-01

Encapsulation of cisplatin anticancer drug into the single walled (10, 0) carbon nanotube and (10, 0) boron-nitride nanotube was investigated by quantum mechanical calculations and Monte Carlo Simulation in aqueous solution. Solvation free energies and complexation free energies of the cisplatin@ carbon nanotube and cisplatin@ boron-nitride nanotube complexes was determined as well as radial distribution functions of entitled compounds. Solvation free energies of cisplatin@ carbon nanotube and cisplatin@ boron-nitride nanotube were − 4.128 kcal mol"−"1 and − 2457.124 kcal mol"−"1 respectively. The results showed that cisplatin@ boron-nitride nanotube was more soluble species in water. In addition electrostatic contribution of the interaction of boron- nitride nanotube complex and solvent was − 281.937 kcal mol"−"1 which really more than Van der Waals and so the electrostatic interactions play a distinctive role in the solvation free energies of boron- nitride nanotube compounds. On the other hand electrostatic part of the interaction of carbon nanotube complex and solvent were almost the same as Van der Waals contribution. Complexation free energies were also computed to study the stability of related structures and the free energies were negative (− 374.082 and − 245.766 kcal mol"−"1) which confirmed encapsulation of drug into abovementioned nanotubes. However, boron-nitride nanotubes were more appropriate for encapsulation due to their larger solubility in aqueous solution. - Highlights: • Solubility of cisplatin@ boron-nitride nanotube is larger than cisplatin@ carbon nanotube. • Boron- nitride nanotube complexes have larger electrostatic contribution in solvation free energy. • Complexation free energies confirm encapsulation of drug into the nanotubes in aqueous solution. • Boron- nitride nanotubes are appropriate drug delivery systems compared with carbon nanotubes.

3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

International Nuclear Information System (INIS)

Choudhry, Nakul; Riyas, A.; Devan, K.; Mohanakrishnan, P.

2013-01-01

Highlights: ► Enhanced burnup potential of existing prototype fast breeder reactor core is studied. ► By increasing the Pu enrichment, fuel burnup can be increased in existing PFBR core. ► Enhanced burnup increase economy and reduce load of fuel fabrication and reprocessing. ► Beginning of life reactivity is suppressed by increasing the number of diluents. ► Absorber rod worth requirements can be achieved by increasing 10 B enrichment. -- Abstract: Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10 B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs

Burnup credit in Spain

International Nuclear Information System (INIS)

Conde, J.M.; Recio, M.

2001-01-01

The status of development of burnup credit for criticality safety analyses in Spain is described in this paper. Ongoing activities in the country in this field, both national and international, are resumed. Burnup credit is currently being applied to wet storage of PWR fuel, and credit to integral burnable absorbers is given for BWR fuel storage. It is envisaged to apply burnup credit techniques to the new generation of transport casks now in the design phase. The analysis methodologies submitted for the analyses of PWR and BWR fuel wet storage are outlined. Analytical activities in the country are described, as well as international collaborations in this field. Perspectives for future research and development of new applications are finally resumed. (author)

Core burn-up calculation method of JRR-3

International Nuclear Information System (INIS)

Kato, Tomoaki; Yamashita, Kiyonobu

2007-01-01

SRAC code system is utilized for core burn-up calculation of JRR-3. SRAC code system includes calculation modules such as PIJ, PIJBURN, ANISN and CITATION for making effective cross section and calculation modules such as COREBN and HIST for core burn-up calculation. As for calculation method for JRR-3, PIJBURN (Cell burn-up calculation module) is used for making effective cross section of fuel region at each burn-up step. PIJ, ANISN and CITATION are used for making effective cross section of non-fuel region. COREBN and HIST is used for core burn-up calculation and fuel management. This paper presents details of NRR-3 core burn-up calculation. FNCA Participating countries are expected to carry out core burn-up calculation of domestic research reactor by SRAC code system by utilizing the information of this paper. (author)

Technological and licensing challenges for high burnup fuel

International Nuclear Information System (INIS)

Gross, H.; Urban, P.; Fenzlein, C.

2002-01-01

Deregulation of electricity markets is driving electricity prices downward as well in the U.S. as in Europe. As a consequence high burnup fuel will be demanded by utilities using either the storage or the reprocessing option. At a minimum, burnups consistent with the current political enrichment limit of 5 w/o will be required for both markets.Significant progress has been achieved in the past by Siemens in meeting the demands of utilities for increased fuel burnup. The technological challenges posed by the increased burnup are mainly related to the corrosion and hydrogen pickup of the clad, the high burnup properties of the fuel and the dimensional changes of the fuel assembly structure. Clad materials with increased corrosion resistance appropriate for high burnup have been developed. The high burnup behaviour of the fuel has been extensively investigated and the decrease of thermal conductivity with burnup, the rim effect of the pellet and the increase of fission gas release with burnup can be described, with good accuracy, in fuel rod computer codes. Advanced statistical design methods have been developed and introduced. Materials with increased corrosion resistance are also helpful controlling the dimensional changes of the fuel assembly structure. In summary, most of the questions about the fuel operational behaviour and reliability in the high burnup range have been solved - some of them are still in the process of verification - or the solutions are visible. This fact is largely acknowledged by regulators too. The main licensing challenges for high burnup fuel are currently seen for accident condition analyses, especially for RIA and LOCA. (author)

Integrated burnup calculation code system SWAT

International Nuclear Information System (INIS)

Suyama, Kenya; Hirakawa, Naohiro; Iwasaki, Tomohiko.

1997-11-01

SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user's manual of SWAT. (author)

Burnup Measurement of Spent Fuel Assembly by CZT-based Gamma-ray Spectroscopy for Input Nuclear Material Accountancy of Pyroprocessing

International Nuclear Information System (INIS)

Seo, Hee; Oh, Jong-Myeong; Shin, Hee-Sung; Kim, Ho-Dong; Lee, Seung-Kyu; Park, Se-Hwan

2013-06-01

Input nuclear material accountancy is crucial for a pyroprocessing facility safeguards. Until a direct Pu measurement technique is established, an indirect method based on code calculations with burnup measurement and neutron counting for 244 Cm could be a practical option. Burnup can be determined by destructive analysis (DA) for final dispositive accuracy or by nondestructive assay (NDA) for near-real time accountancy. In the present study, an underwater burnup measurement system based on gamma-ray spectroscopy with the CZT detector was developed and tested on a spent fuel assembly. Burnup was determined according to the 134 Cs/ 137 Cs activity ratio with efficiency correction by Geant4 Monte Carlo simulations. The activity ratio as a function of burnup was obtained by ORIGEN calculations. The measured burnup error was 8.6%, which was within the measurement uncertainty. It is expected that the underwater burnup measurement system could fulfill an important role as a means of near-real time accountancy at a future pyroprocessing facility. (authors)

Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

International Nuclear Information System (INIS)

Kim, Jung Suk; Jeon, Young Shin; Park, Soon Dal; Ha, Yeong Keong; Song, Kyu Seok

2015-01-01

The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional 235 U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using 233 U, 242 Pu, 150 Nd, and 133 Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code

'CANDLE' burnup regime after LWR regime

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Nagata, Akito

2008-01-01

CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium. (author)

Molecular dynamics simulation of deformation twin in rocksalt vanadium nitride

International Nuclear Information System (INIS)

Fu, Tao; Peng, Xianghe; Zhao, Yinbo; Li, Tengfei; Li, Qibin; Wang, Zhongchang

2016-01-01

We perform molecular dynamics simulation of nano-indentation with a cylindrical indenter to investigate the formation mechanism of deformation twin in vanadium nitride (VN) with a rocksalt structure. We find that the deformation twins occur during the loading stage, and subsequently conduct a systematic analysis of nucleation, propagation and thickening of a deformation twin. We find that the nucleation of a partial dislocation and its propagation to form a stacking fault are premise of deformation twin formation. The sequential nucleation and propagation of partial dislocation on adjacent parallel {111} planes are found to cause the thickening of the deformation twin. Moreover, the deformation twins can exist in VN at room temperature. - Highlights: • MD simulations of indentation are performed to study the deformation twin in VN. • The deformation twins can occur in VN during the loading stage. • The nucleation, propagation and thickening of a deformation twin are analyzed. • The deformation twins can exist in VN at room temperature.

Simulation of triton burn-up in JET plasmas

Energy Technology Data Exchange (ETDEWEB)

Loughlin, M J; Balet, B; Jarvis, O N; Stubberfield, P M [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

1994-07-01

This paper presents the first triton burn-up calculations for JET plasmas using the transport code TRANSP. Four hot ion H-mode deuterium plasmas are studied. For these discharges, the 2.5 MeV emission rises rapidly and then collapses abruptly. This phenomenon is not fully understood but in each case the collapse phase is associated with a large impurity influx known as the ``carbon bloom``. The peak 14 MeV emission occurs at this time, somewhat later than that of the 2.5 MeV neutron peak. The present results give a clear indication that there are no significant departures from classical slowing down and spatial diffusion for tritons in JET plasmas. (authors). 7 refs., 3 figs., 1 tab.

Increased burnup of fuel elements

International Nuclear Information System (INIS)

Ahlf, J.

1983-01-01

The specialists' group for fuel elements of the Kerntechnische Gesellschaft e.V. held a meeting on ''Increased Burnup of Fuel Elements'' on 9th and 10th of November 1982 at the GKSS Research Center Geesthacht. Most papers dealt with the problems of burnup increase of fuel elements for light water reactors with respect to fuel manufacturing, power plant operation and reprocessing. Review papers were given on the burnup limits for high temperature gas cooled reactors and sodium fast breeder reactors. The meeting ended with a presentation of the technical equipment of the hot laboratory of the GKSS and the programs which are in progress there. (orig.) [de

Evaluation of Gap Conductance Approach for Mid-Burnup Fuel LOCA Analysis

International Nuclear Information System (INIS)

Lee, Joosuk; Woo, Swengwoong

2013-01-01

In this study, therefore, the applicability of gap conductance approach on the mid-burnup fuel in LOCA analysis was estimated in terms of the comparison of PCT distribution method means the fuel rod uncertainty is taken into account by the combination of overall uncertainty parameters of fuel rod altogether by use of a simple random sampling(SRS) technique. There are many uncertainty parameters of fuel rod that can change the PCT during LOCA analysis, and these have been identified by the authors' previous work already. But, for the 'best-estimate' LOCA safety analysis the methodology that dose not use the overall uncertainty parameters altogether but used the gap conductance uncertainty alone has been developed to simulate the overall fuel rod uncertainty, because it can represent many uncertainty parameters. Based on this approach, uncertainty range of gap conductance was prescribed as 0.67∼1.5 in audit calculation methodology on LBLOCA analysis. This uncertainty was derived from experimental data of fresh or low burnup fuel. Meanwhile, recent research work identify that the currently utilized uncertainty range seems to be not enough to encompass the uncertainty of mid-burnup fuel. Instead it has to be changed to 0.5∼2.4 for the mid-burnup fuel(30 MWd/kgU)

Evaluation of Gap Conductance Approach for Mid-Burnup Fuel LOCA Analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, Joosuk; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

In this study, therefore, the applicability of gap conductance approach on the mid-burnup fuel in LOCA analysis was estimated in terms of the comparison of PCT distribution method means the fuel rod uncertainty is taken into account by the combination of overall uncertainty parameters of fuel rod altogether by use of a simple random sampling(SRS) technique. There are many uncertainty parameters of fuel rod that can change the PCT during LOCA analysis, and these have been identified by the authors' previous work already. But, for the 'best-estimate' LOCA safety analysis the methodology that dose not use the overall uncertainty parameters altogether but used the gap conductance uncertainty alone has been developed to simulate the overall fuel rod uncertainty, because it can represent many uncertainty parameters. Based on this approach, uncertainty range of gap conductance was prescribed as 0.67∼1.5 in audit calculation methodology on LBLOCA analysis. This uncertainty was derived from experimental data of fresh or low burnup fuel. Meanwhile, recent research work identify that the currently utilized uncertainty range seems to be not enough to encompass the uncertainty of mid-burnup fuel. Instead it has to be changed to 0.5∼2.4 for the mid-burnup fuel(30 MWd/kgU)

Measurement of burnup in FBR MOX fuel irradiated to high burnup

International Nuclear Information System (INIS)

Koyama, Shin-ichi; Osaka, Masahiko; Sekine, Takashi; Morozumi, Katsufumi; Namekawa, Takashi; Itoh, Masahiko

2003-01-01

The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28% FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector. (author)

VAMPIR - A two-group two-dimensional diffusion computer code for burnup calculation

International Nuclear Information System (INIS)

Zmijarevic, I.; Petrovic, I.

1985-01-01

VAMPIR is a computer code which simulates the burnup within a reactor coe. It computes the neutron flux, power distribution and burnup taking into account spatial variations of temperature and xenon poisoning. Its overall reactor calculation uses diffusion theory with finite differences approximation in X-Y or R-Z geometry. Two-group macroscopic cross section data are prepared by the lattice cell code WIMS-D4 and stored in the library form of multi entry tabulation against the various parameters that significantly affect the physical conditions in the reactor core. herein, the main features of the program are presented. (author)

Advancements in reactor physics modelling methodology of Monte Carlo Burnup Code MCB dedicated to higher simulation fidelity of HTR cores

International Nuclear Information System (INIS)

Cetnar, Jerzy

2014-01-01

The recent development of MCB - Monte Carlo Continuous Energy Burn-up code is directed towards advanced description of modern reactors, including double heterogeneity structures that exist in HTR-s. In this, we exploit the advantages of MCB methodology in integrated approach, where physics, neutronics, burnup, reprocessing, non-stationary process modeling (control rod operation) and refined spatial modeling are carried in a single flow. This approach allows for implementations of advanced statistical options like analysis of error propagation, perturbation in time domain, sensitivity and source convergence analyses. It includes statistical analysis of burnup process, emitted particle collection, thermal-hydraulic coupling, automatic power profile calculations, advanced procedures of burnup step normalization and enhanced post processing capabilities. (author)

Issues for effective implementation of burnup credit

International Nuclear Information System (INIS)

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

2001-01-01

In the United States, burnup credit has been used in the criticality safety evaluation for storage pools at pressurized water reactors (PWRs) and considerable work has been performed to lay the foundation for use of burnup credit in dry storage and transport cask applications and permanent disposal applications. Many of the technical issues related to the basic physics phenomena and parameters of importance are similar in each of these applications. However, the nuclear fuel cycle in the United States has never been fully integrated and the implementation of burnup credit to each of these applications is dependent somewhat on the specific safety bases developed over the history of each operational area. This paper will briefly review the implementation status of burnup credit for each application area and explore some of the remaining issues associated with effective implementation of burnup credit. (author)

Molecular-dynamics simulation of defect formation energy in boron nitride nanotubes

International Nuclear Information System (INIS)

Moon, W.H.; Hwang, H.J.

2004-01-01

We investigate the defect formation energy of boron nitride nanotubes (BNNTs) using molecular dynamics simulation. Although the defect with tetragon-octagon pairs (TOP) is favored in the flat BNNTs cap, BN clusters, and the growth of BNNTs, the formation energy of the TOP defect is significantly higher than that of the pentagon-heptagon pairs (PHP) defect in BNNTs. The PHP defect reduces the effect of the structural distortion caused by the TOP defect, in spite of homoelemental bonds. The instability of the TOP defect generates the structural transformation into BNNTs with no defect at about 1500 K. This mechanism shows that the TOP defect is less favored in case of BNNTs

Investigation and basic evaluation for ultra-high burnup fuel cladding material

International Nuclear Information System (INIS)

Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi

2001-03-01

In ultra-high burnup of the power reactor, it is an essential problem to develop the cladding with excellent durability. First, development history and approach of the safety assessment of Zircaloy for the high burnup fuel were summarized in the report. Second, the basic evaluation and investigation were carried out on the material with high practicability in order to select the candidate materials for the ultra-high burnup fuel. In addition, the basic research on modification technology of the cladding surface was carried out from the viewpoint of the addition of safety margin as a cladding. From the development history of the zirconium alloy including the Zircaloy, it is hard to estimate the results of in-pile test from those of the conventional corrosion test (out-pile test). Therefore, the development of the new testing technology that can simulate the actual environment and the elucidation of the corrosion-controlling factor of the cladding are desired. In cases of RIA (Reactivity Initiated Accident) and LOCA (Loss of Coolant Accident), it seems that the loss of ductility in zirconium alloys under heavy irradiation and boiling of high temperature water restricts the extension of fuel burnup. From preliminary evaluation on the high corrosion-resistance materials (austenitic stainless steel, iron or nickel base superalloys, titanium alloy, niobium alloy, vanadium alloy and ferritic stainless steel), stabilized austenitic stainless steels with a capability of future improvement and high-purity niobium alloys with a expectation of the good corrosion resistance were selected as candidate materials of ultra-high burnup cladding. (author)

Systemization of burnup sensitivity analysis code

International Nuclear Information System (INIS)

Tatsumi, Masahiro; Hyoudou, Hideaki

2004-02-01

To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this

Phenomena and parameters important to burnup credit

International Nuclear Information System (INIS)

Parks, C.V.; Dehart, M.D.; Wagner, J.C.

2001-01-01

Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water- reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the United States and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given. (author)

Thermodynamic Studies of Decane on Boron Nitride and Graphite Substrates Using Synchrotron Radiation and Molecular Dynamics Simulations

Science.gov (United States)

Strange, Nicholas; Arnold, Thomas; Forster, Matthew; Parker, Julia; Larese, J. Z.; Diamond Light Source Collaboration; University of Tennessee Team

2014-03-01

Hexagonal boron nitride (hBN) has a lattice structure similar to that of graphite with a slightly larger lattice parameter in the basal plane. This, among other properties, makes it an excellent substrate in place of graphite, eliciting some important differences. This work is part of a larger effort to examine the interaction of alkanes with magnesium oxide, graphite, and boron nitride surfaces. In our current presentation, we will discuss the interaction of decane with these surfaces. Decane exhibits a fully commensurate structure on graphite and hBN at monolayer coverages. In this particular experiment, we have examined the monolayer structure of decane adsorbed on the basal plane of hBN using synchrotron x-ray radiation at Diamond Light Source. Additionally, we have examined the system experimentally with volumetric isotherms as well as computationally using molecular dynamics simulations. The volumetric isotherms allow us to calculate properties which provide important information about the adsorbate's interaction with not only neighboring molecules, but also the interaction with the adsorbent boron nitride.

Platinum group metal nitrides and carbides: synthesis, properties and simulation

International Nuclear Information System (INIS)

Ivanovskii, Alexander L

2009-01-01

Experimental and theoretical data on new compounds, nitrides and carbides of the platinum group 4d and 5d metals (ruthenium, rhodium, palladium, osmium, iridium, platinum), published over the past five years are summarized. The extreme mechanical properties of platinoid nitrides and carbides, i.e., their high strength and low compressibility, are noted. The prospects of further studies and the scope of application of these compounds are discussed.

Study on the sensitivity of Self-Powered Neutron Detectors (SPND) and its change due to burn-up

International Nuclear Information System (INIS)

Cho, Gyuseong; Lee, Wanno; Yoon, Jeong-Hyoun.

1996-01-01

Self-Powered Neutron Detectors (SPND) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. While they have several advantages such as small size, low cost, and relatively simple electronics required in conjunction with its usage, it has some intrinsic problems of the low level of output current, a slow response time, the rapid change of sensitivity which makes it difficult to use for a long term. In this paper, Monte Carlo simulation was accomplished to calculate the escape probability as a function of the birth position for the typical geometry of rhodium-based SPNDs. Using the simulation result, the burn-up profile of rhodium number density and the neutron sensitivity is calculated as a function of burn-up time in the reactor. The sensitivity of the SPND decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long-term usage. (author)

PWR AXIAL BURNUP PROFILE ANALYSIS

International Nuclear Information System (INIS)

J.M. Acaglione

2003-01-01

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

Burnup calculation for a tokamak commercial hybrid reactor

International Nuclear Information System (INIS)

Feng Kaiming; Xie Zhongyou

1990-08-01

A computer code ISOGEN-III and its associated data library BULIB have been developed for fusion-fission hybrid reactor burnup calculations. These are used to calcuate burnup of a tokamak commercial hybrid reactor. The code and library are introduced briefly, and burnup calculation results are given

Burnup credit activities being conducted in the United States

International Nuclear Information System (INIS)

Lake, W.

1998-01-01

The paper describes burnup credit activities being conducted in the U.S. where burnup credit is either being used or being planned to be used for storage, transport, and disposal of spent nuclear fuel. Currently approved uses of burnup credit are for wet storage of PWR fuel. For dry storage of spent PWR fuel, burnup credit is used to supplement a principle of moderator exclusion. These storage applications have been pursued by the private sector. The Department of Energy (DOE) which is an organization of the U.S. Federal government is seeking approval for burnup credit for transport and disposal applications. For transport of spent fuel, regulatory review of an actinide-only PWR burnup credit method is now being conducted. A request by DOE for regulatory review of actinide and fission product burnup credit for disposal of spent BWR and PWR fuel is scheduled to occur in 1998. (author)

Electrochemical properties of lanthanum nitride with calcium nitride additions

International Nuclear Information System (INIS)

Lesunova, R.P.; Fishman, L.S.

1986-01-01

This paper reports on the electrochemical properties of lanthanum nitride with calcium nitride added. The lanthanum nitride was obtained by nitriding metallic lanthanum at 870 K in an ammonia stream. The product contained Cl, Pr, Nd, Sm, Fe, Ca, Cu, Mo, Mg, Al, Si, and Be. The calcium nitride was obtained by nitriding metallic calcium in a nitrogen stream. The conductivity on the LaN/C 3 N 2 system components are shown as a function of temperature. A table shows the solid solutions to be virtually electronic conductors and the lanthanum nitride a mixed conductor

COGEMA/TRANSNUCLEAIRE's experience with burnup credit

International Nuclear Information System (INIS)

Chanzy, Y.; Guillou, E.

1998-01-01

Facing a continuous increase in the fuel enrichments, COGEMA and TRANSNUCLEAIRE have implemented step by step a burnup credit programme to improve the capacity of their equipment without major physical modification. Many authorizations have been granted by the French competent authority in wet storage, reprocessing and transport since 1981. As concerns transport, numerous authorizations have been validated by foreign competent authorities. Up to now, those authorizations are restricted to PWR Fuel type assemblies made of enriched uranium. The characterization of the irradiated fuel and the reactivity of the systems are evaluated by calculations performed with well qualified French codes developed by the CEA (French Atomic Energy Commission): CESAR as a depletion code and APPOLO-MORET as a criticality code. The authorizations are based on the assurance that the burnup considered is met on the least irradiated part of the fuel assemblies. Besides, the most reactive configuration is calculated and the burnup credit is restricted to major actinides only. This conservative approach allows not to take credit for any axial profile. On the operational side, the procedures have been reevaluated to avoid misloadings and a burnup verification is made before transport, storage and reprocessing. Depending on the level of burnup credit, it consists of a qualitative (go/no-go) verification or of a quantitative measurement. Thus the use of burnup credit is now a common practice in France and Germany and new improvements are still in progress: extended qualifications of the codes are made to enable the use of six selected fission products in the criticality evaluations. (author)

Impacts of burnup-dependent swelling of metallic fuel on the performance of a compact breed-and-burn fast reactor

Energy Technology Data Exchange (ETDEWEB)

Hartanto, Donny; Heo, Woong; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

2016-04-15

The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

Automated generation of burnup chain for reactor analysis applications

International Nuclear Information System (INIS)

Tran, Viet-Phu; Tran, Hoai-Nam; Yamamoto, Akio; Endo, Tomohiro

2017-01-01

This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO_2 and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

Automated generation of burnup chain for reactor analysis applications

Energy Technology Data Exchange (ETDEWEB)

Tran, Viet-Phu [VINATOM, Hanoi (Viet Nam). Inst. for Nuclear Science and Technology; Tran, Hoai-Nam [Duy Tan Univ., Da Nang (Viet Nam). Inst. of Research and Development; Yamamoto, Akio; Endo, Tomohiro [Nagoya Univ., Nagoya-shi (Japan). Dept. of Materials, Physics and Energy Engineering

2017-05-15

This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO{sub 2} and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

Systemization of burnup sensitivity analysis code. 2

International Nuclear Information System (INIS)

Tatsumi, Masahiro; Hyoudou, Hideaki

2005-02-01

Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For

Benchmarking burnup reconstruction methods for dynamically operated research reactors

Energy Technology Data Exchange (ETDEWEB)

Sternat, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Charlton, William S. [Univ. of Nebraska, Lincoln, NE (United States). National Strategic Research Institute; Nichols, Theodore F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-03-01

The burnup of an HEU fueled dynamically operated research reactor, the Oak Ridge Research Reactor, was experimentally reconstructed using two different analytic methodologies and a suite of signature isotopes to evaluate techniques for estimating burnup for research reactor fuel. The methods studied include using individual signature isotopes and the complete mass spectrometry spectrum to recover the sample’s burnup. The individual, or sets of, isotopes include ¹⁴⁸Nd, ¹³⁷Cs+¹³⁷Ba, ¹³⁹La, and ¹⁴⁵Nd+¹⁴⁶Nd. The storage documentation from the analyzed fuel material provided two different measures of burnup: burnup percentage and the total power generated from the assembly in MWd. When normalized to conventional units, these two references differed by 7.8% (395.42GWd/MTHM and 426.27GWd/MTHM) in the resulting burnup for the spent fuel element used in the benchmark. Among all methods being evaluated, the results were within 11.3% of either reference burnup. The results were mixed in closeness to both reference burnups; however, consistent results were achieved from all three experimental samples.

Burnup performance of rock-like oxide (ROX) fuel in small pebble bed reactor with accumulative fuel loading scheme

International Nuclear Information System (INIS)

Simanullang, Irwan Liapto; Obara, Toru

2017-01-01

Highlights: • Burnup performance using ROX fuel in PBR with accumulative fuel loading scheme was analyzed. • Initial excess reactivity was suppressed by reducing 235 U enrichment in the startup condition. • Negative temperature coefficient was achieved in all condition of PBR with accumulative fuel loading scheme using ROX fuel. • Core lifetime of PBR with accumulative fuel loading scheme using ROX fuel was shorter than with UO 2 fuel. • In PBR with accumulative fuel loading scheme using ROX fuel, achieved discharged burnup can be as high as that for UO 2 fuel. - Abstract: The Japan Atomic Energy Agency (JAEA) has proposed rock-like oxide (ROX) fuel as a new, once-through type fuel concept. Here, burnup performance using ROX fuel was simulated in a pebble bed reactor with an accumulative fuel loading scheme. The MVP-BURN code was used to simulate the burnup calculation. Fuel of 5 g-HM/pebble with 20% 235 U enrichment was selected as the optimum composition. Discharged burnup could reach up to 218 GWd/t, with a core lifetime of about 8.4 years. However, high excess reactivity occurred in the initial condition. Initial fuel enrichment was therefore reduced from 20% to 4.65% to counter the initial excess reactivity. The operation period was reduced by the decrease of initial fuel enrichment, but the maximum discharged burnup was 198 GWd/t. Burnup performance of ROX fuel in this reactor concept was compared with that of UO 2 fuel obtained previously. Discharged burnup for ROX fuel in the PBR with an accumulative fuel loading scheme was as high as UO 2 fuel. Maximum power density could be lowered by introducing ROX fuel compared to UO 2 fuel. However, PBR core lifetime was shorter with ROX fuel than with UO 2 fuel. A negative temperature coefficient was achieved for both UO 2 and ROX fuels throughout the operation period.

The build-up and characterization of nuclear burn-up wave in a fast ...

Indian Academy of Sciences (India)

K V Anoop

2018-02-07

Feb 7, 2018 ... evaluating the quality of the wave by the researchers working in the field of nuclear burn-up wave build-up and propagation. Keywords. ... However, there are concerns relating to the nuclear safety, ... Simulation studies have.

Burnup credit activities in the United States

International Nuclear Information System (INIS)

Lake, W.H.; Thomas, D.A.; Doering, T.W.

2001-01-01

This report covers progress in burnup credit activities that have occurred in the United States of America (USA) since the International Atomic Energy Agency's (IAEA's) Advisory Group Meeting (AGM) on Burnup Credit was convened in October 1997. The Proceeding of the AGM were issued in April 1998 (IAEA-TECDOC-1013, April 1998). The three applications of the use of burnup credit that are discussed in this report are spent fuel storage, spent fuel transportation, and spent fuel disposal. (author)

K-infinite trends with burnup, enrichment, and cooling time for BWR fuel assemblies

International Nuclear Information System (INIS)

Broadhead, B.L.

1998-08-01

This report documents the work performed by ORNL for the Yucca Mountain project (YMP) M and O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k inf values for infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various burnup/enrichment and cooling-time combinations. These scenarios simulate expected limiting criticality loading conditions (for a given assembly type) for drift emplacements in a repository. Upon consultation with the YMP staff, a Quad Cities BWR fuel assembly was selected as a baseline assembly. This design consists of seven axial enrichment zones, three of which contain natural uranium oxide. No attempt was made to find a bounding or even typical assembly design due to the wide variety in fuel assembly designs necessary for consideration. The current work concentrates on establishing a baseline analysis, along with a small number of sensitivity studies which can be expected later if desired. As a result of similar studies of this nature, several effects are known to be important in the determination of the final k inf for spent fuel in a cask-like geometry. For a given enrichment there is an optimal burnup: for lower burnups, excess energy (and corresponding excess reactivity) is present in the fuel assembly; for larger burnups, the assembly is overburned and essentially driven by neighboring fuel assemblies. The majority of the burnup/enrichment scenarios included in this study were for some near-optimum burnup/enrichment combinations as determined from Energy Information Administration (EIA) data. Several calculations were performed for under- and over-burned fuel to show these effects

PENBURN - A 3-D Zone-Based Depletion/Burnup Solver

International Nuclear Information System (INIS)

Manalo, Kevin; Plower, Thomas; Rowe, Mireille; Mock, Travis; Sjoden, Glenn E.

2008-01-01

PENBURN (Parallel Environment Burnup) is a general depletion/burnup solver which, when provided with zone-based reaction rates, computes time-dependent isotope concentrations for a set of actinides and fission products. Burnup analysis in PENBURN is performed with a direct Bateman-solver chain solution technique. Specifically, in tandem with PENBURN is the use of PENTRAN, a parallel multi-group anisotropic Sn code for 3-D Cartesian geometries. In PENBURN, the linear chain method is actively used to solve individual isotope chains which are then fully attributed by the burnup code to yield integrated isotope concentrations for each nuclide specified. Included with the discussion of code features, a single PWR fuel pin calculation with the burnup code is performed and detailed with a benchmark comparison to PIE (Post-Irradiation Examination) data within the SFCOMPO (Spent Fuel Composition / NEA) database, and also with burnup codes in SCALE5.1. Conclusions within the paper detail, in PENBURN, the accuracy of major actinides, flux profile behavior as a function of burnup, and criticality calculations for the PWR fuel pin model. (authors)

A Monte Carlo burnup code linking MCNP and REBUS

International Nuclear Information System (INIS)

Hanan, N. A.

1998-01-01

The REBUS-3 burnup code, used in the ANL RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult burnup analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented

Implementation of burnup credit in spent fuel management systems

International Nuclear Information System (INIS)

Dyck, H.P.

2001-01-01

Improved calculational methods allow one to take credit for the reactivity reduction associated with fuel burnup. This means reducing the analysis conservatism while maintaining an adequate safety margin. The motivation for using burnup credit in criticality safety applications is based on economic considerations and additional benefits contributing to public health and safety and resource conservation. Interest in the implementation of burnup credit has been shown by many countries. In 1997, the International Atomic Energy Agency (IAEA) started a task to monitor the implementation of burnup credit in spent fuel management systems, to provide a forum to exchange information, to discuss the matter and to gather and disseminate information on the status of national practices of burnup credit implementation in the Member States. The task addresses current and future aspects of burnup credit. This task was continued during the following years. (author)

Probabilistic assessment of dry transport with burnup credit

International Nuclear Information System (INIS)

Lake, W.H.

2003-01-01

The general concept of probabilistic analysis and its application to the use of burnup credit in spent fuel transport is explored. Discussion of the probabilistic analysis method is presented. The concepts of risk and its perception are introduced, and models are suggested for performing probability and risk estimates. The general probabilistic models are used for evaluating the application of burnup credit for dry spent nuclear fuel transport. Two basic cases are considered. The first addresses the question of the relative likelihood of exceeding an established criticality safety limit with and without burnup credit. The second examines the effect of using burnup credit on the overall risk for dry spent fuel transport. Using reasoned arguments and related failure probability and consequence data analysis is performed to estimate the risks of using burnup credit for dry transport of spent nuclear fuel. (author)

Fast reactor cycle calculation routine using a 3D-simulator and investigation of new burnup stategies for pressurized water reactors

International Nuclear Information System (INIS)

Li Yulun.

1987-03-01

Three-dimensional calculations of the longtime behaviour of PWR can be done in short computing times with satisfactory accuracy for power and burn-up distributions. This has been proved by comparison with operational data of Biblis-B. Various possibilities are investigated to increase the discharge burn-up and to improve the utilization of uranium. In view of the increase of discharge burn-up due to enhanced cycle number (decreased batch size) and decreased neutron leakage these new strategies are intensively studied in the conventional fuel management scheme (Out-in) and in the low leakage fuel management scheme (In-Out). By a conventional fuel management scheme with four cycle operation and a low leakage fuel management scheme with three cycle operation an attractive increase of discharge burn-up to about 40% can be achieved by an increase in the reload enrichment to 4%. (orig.) [de

Superconducting structure with layers of niobium nitride and aluminum nitride

International Nuclear Information System (INIS)

Murduck, J.M.; Lepetre, Y.J.; Schuller, I.K.; Ketterson, J.B.

1989-01-01

A superconducting structure is formed by depositing alternate layers of aluminum nitride and niobium nitride on a substrate. Deposition methods include dc magnetron reactive sputtering, rf magnetron reactive sputtering, thin-film diffusion, chemical vapor deposition, and ion-beam deposition. Structures have been built with layers of niobium nitride and aluminum nitride having thicknesses in a range of 20 to 350 Angstroms. Best results have been achieved with films of niobium nitride deposited to a thickness of approximately 70 Angstroms and aluminum nitride deposited to a thickness of approximately 20 Angstroms. Such films of niobium nitride separated by a single layer of aluminum nitride are useful in forming Josephson junctions. Structures of 30 or more alternating layers of niobium nitride and aluminum nitride are useful when deposited on fixed substrates or flexible strips to form bulk superconductors for carrying electric current. They are also adaptable as voltage-controlled microwave energy sources. 8 figs

High burnup MOX fuel assembly

International Nuclear Information System (INIS)

Blanpain, P.; Brunel, L.

1999-01-01

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

Burnup credit applications in a high-capacity truck cask

International Nuclear Information System (INIS)

Boshoven, J.K.

1993-01-01

The use of burnup credit in the criticality safety analysis of the GA-4 Cask increases the cask's capacity from three spent fuel assemblies to four, resulting in reduced public and occupational risk and reduced life cycle costs. GA's criticality calculations for burnup credit, including the associated uncertainties and analytical bias, establish the minimum burnup required as a function of initial enrichment to maintain K eff ≤ 0.95 under any conceivable condition. The minimum burnup requirement as a function of initial enrichment has been determined to be 15,000 MWd/MTU for 3.5 wt% U-235 fuel, 20,000 MWd/MTU for 4.0 wt% U-235 fuel and 25,000 MWd/MTU for 4.5 wt% U-235 fuel. The minimum burnup requirement as a function of enrichment is well below the typical burnup levels seen in the current and projected spent fuel inventory. (J.P.N.)

Assesment of advanced step models for steady state Monte Carlo burnup calculations in application to prismatic HTGR

Directory of Open Access Journals (Sweden)

Kępisty Grzegorz

2015-09-01

Full Text Available In this paper, we compare the methodology of different time-step models in the context of Monte Carlo burnup calculations for nuclear reactors. We discuss the differences between staircase step model, slope model, bridge scheme and stochastic implicit Euler method proposed in literature. We focus on the spatial stability of depletion procedure and put additional emphasis on the problem of normalization of neutron source strength. Considered methodology has been implemented in our continuous energy Monte Carlo burnup code (MCB5. The burnup simulations have been performed using the simplified high temperature gas-cooled reactor (HTGR system with and without modeling of control rod withdrawal. Useful conclusions have been formulated on the basis of results.

Application of reactivity method to MTR fuel burn-up measurement

International Nuclear Information System (INIS)

Zuniga, A.; Ravnik, M.; Cuya, R.

2001-01-01

Fuel element burn-up has been measured for the first time by reactivity method in a MTR reactor. The measurement was performed in RP-10 reactor of Peruvian Institute for Nuclear Energy (IPEN) in Lima. It is a pool type 10MW material testing reactor using standard 20% enriched uranium plate type fuel elements. A fresh element and an element with well defined burn-up were selected as reference elements. Several elements in the core were selected for burn-up measurement. Each of them was replaced in its original position by both reference elements. Change in excess reactivity was measured using control rod calibration curve. The burn-up reactivity worth of fuel elements was plotted as a function of their calculated burnup. Corrected burn-up values of the measured fuel elements were calculated using the fitting function at experimental reactivity for all elements. Good agreement between measured and calculated burn-up values was observed indicating that the reactivity method can be successfully applied also to MTR fuel element burn-up determination.(author)

Application of Candle burnup to small fast reactor

International Nuclear Information System (INIS)

Sekimoto, H.; Satoshi, T.

2004-01-01

A new reactor burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. An equilibrium state was obtained for a large fast reactor (core radius is 2 m and reflector thickness is 0.5 m) successfully by using a newly developed direct analysis code. However, it is difficult to apply this burnup strategy to small reactors, since its neutron leakage becomes large and neutron economy becomes worse. Fuel enrichment should be increased in order to sustain the criticality. However, higher enrichment of fresh fuel makes the CANDLE burnup difficult. We try to find some small reactor designs, which can realize the CANDLE burnup. We have successfully find a design, which is not the CANDLE burnup in the strict meaning, but satisfies qualitatively its characteristics mentioned at the top of this abstract. In the final paper, the general description of CANDLE burnup and some results on the obtained small fast reactor design are presented.(author)

Burn-up measurements coupling gamma spectrometry and neutron measurement

Energy Technology Data Exchange (ETDEWEB)

Toubon, H.; Pin, P. [AREVA/CANBERRA, 1 rue des Herons, 78182 St Quentin-en-Yvelines Cedex (France); Lebrun, A. [IAEA, Wagramer Strasse 5, PO Box 100, Vienna (Austria); Oriol, L.; Saurel, N. [CEA Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Gain, T. [AREVA/COGEMA Reprocessing Business Unit, La Hague, 50444 Beaumont Hague Cedex (France)

2006-07-01

The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

Burn-up measurements coupling gamma spectrometry and neutron measurement

International Nuclear Information System (INIS)

Toubon, H.; Pin, P.; Lebrun, A.; Oriol, L.; Saurel, N.; Gain, T.

2006-01-01

The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

An empirical formulation to describe the evolution of the high burnup structure

Energy Technology Data Exchange (ETDEWEB)

Lemes, Martín; Soba, Alejandro; Denis, Alicia

2015-01-15

In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

The oxidation of titanium nitride- and silicon nitride-coated stainless steel in carbon dioxide environments

International Nuclear Information System (INIS)

Mitchell, D.R.G.; Stott, F.H.

1992-01-01

A study has been undertaken into the effects of thin titanium nitride and silicon nitride coatings, deposited by physical vapour deposition and chemical vapour deposition processes, on the oxidation resistance of 321 stainless steel in a simulated advanced gas-cooled reactor carbon dioxide environment for long periods at 550 o C and 700 o C under thermal-cycling conditions. The uncoated steel contains sufficient chromium to develop a slow-growing chromium-rich oxide layer at these temperatures, particularly if the surfaces have been machine-abraded. Failure of this layer in service allows formation of less protective iron oxide-rich scales. The presence of a thin (3-4 μm) titanium nitride coating is not very effective in increasing the oxidation resistance since the ensuing titanium oxide scale is not a good barrier to diffusion. Even at 550 o C, iron oxide-rich nodules are able to develop following relatively rapid oxidation and breakdown of the coating. At 700 o C, the coated specimens oxidize at relatively similar rates to the uncoated steel. A thin silicon nitride coating gives improved oxidation resistance, with both the coating and its slow-growing oxide being relatively electrically insulating. The particular silicon nitride coating studied here was susceptible to spallation on thermal cycling, due to an inherently weak coating/substrate interface. (Author)

Burnup calculations using Monte Carlo method

International Nuclear Information System (INIS)

Ghosh, Biplab; Degweker, S.B.

2009-01-01

In the recent years, interest in burnup calculations using Monte Carlo methods has gained momentum. Previous burn up codes have used multigroup transport theory based calculations followed by diffusion theory based core calculations for the neutronic portion of codes. The transport theory methods invariably make approximations with regard to treatment of the energy and angle variables involved in scattering, besides approximations related to geometry simplification. Cell homogenisation to produce diffusion, theory parameters adds to these approximations. Moreover, while diffusion theory works for most reactors, it does not produce accurate results in systems that have strong gradients, strong absorbers or large voids. Also, diffusion theory codes are geometry limited (rectangular, hexagonal, cylindrical, and spherical coordinates). Monte Carlo methods are ideal to solve very heterogeneous reactors and/or lattices/assemblies in which considerable burnable poisons are used. The key feature of this approach is that Monte Carlo methods permit essentially 'exact' modeling of all geometrical detail, without resort to ene and spatial homogenization of neutron cross sections. Monte Carlo method would also be better for in Accelerator Driven Systems (ADS) which could have strong gradients due to the external source and a sub-critical assembly. To meet the demand for an accurate burnup code, we have developed a Monte Carlo burnup calculation code system in which Monte Carlo neutron transport code is coupled with a versatile code (McBurn) for calculating the buildup and decay of nuclides in nuclear materials. McBurn is developed from scratch by the authors. In this article we will discuss our effort in developing the continuous energy Monte Carlo burn-up code, McBurn. McBurn is intended for entire reactor core as well as for unit cells and assemblies. Generally, McBurn can do burnup of any geometrical system which can be handled by the underlying Monte Carlo transport code

Effect of burnup on the response of stainless steel-clad mixed-oxide fuels to simulated thermal transients

International Nuclear Information System (INIS)

Fenske, G.R.; Badyopadhyay, G.

1981-01-01

Direct electrical heating experiments were performed on irradiated fuel to study the fuel and cladding response as a function of burnup during a slow thermal transient. The results indicated that the nature and extent of the fuel and cladding behavior depended on the quantity of fission gas retained in the fuel. Fission-gas-driven fuel ejection occurred as the molten cladding flowed down the stack exposing bare, radially unrestrained fuel. The fuel dispersion occurred in the absence of molten fuel and the amount of fuel ejected increased with increasing burnup. 31 refs

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)

Disposal criticality analysis methodology's principal isotope burnup credit

International Nuclear Information System (INIS)

Doering, T.W.; Thomas, D.A.

2001-01-01

This paper presents the burnup credit aspects of the United States Department of Energy Yucca Mountain Project's methodology for performing criticality analyses for commercial light-water-reactor fuel. The disposal burnup credit methodology uses a 'principal isotope' model, which takes credit for the reduced reactivity associated with the build-up of the primary principal actinides and fission products in irradiated fuel. Burnup credit is important to the disposal criticality analysis methodology and to the design of commercial fuel waste packages. The burnup credit methodology developed for disposal of irradiated commercial nuclear fuel can also be applied to storage and transportation of irradiated commercial nuclear fuel. For all applications a series of loading curves are developed using a best estimate methodology and depending on the application, an additional administrative safety margin may be applied. The burnup credit methodology better represents the 'true' reactivity of the irradiated fuel configuration, and hence the real safety margin, than do evaluations using the 'fresh fuel' assumption. (author)

Fission-product burnup chain model for research reactor application

Energy Technology Data Exchange (ETDEWEB)

Kim, Jung Do; Gil, Choong Sup; Lee, Jong Tai [Korea Atomic Energy Research Inst., Daeduk (Republic of Korea)

1990-12-01

A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the preudo-element and the preudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.(Author).

Features of fuel performance at high fuel burnups

International Nuclear Information System (INIS)

Proselkov, V.N.; Scheglov, A.S.; Smirnov, A.V.; Smirnov, V.P.

2001-01-01

Some features of fuel behavior at high fuel burnups, in particular, initiation and development of rim-layer, increase in the rate of fission gas release from the fuel and increase in the inner gas pressure in the fuel rod are briefly described. Basing on the analysis of the data of post-irradiation examinations of fuel rods of WWER-440 working FA and CR fuel followers, that have been operated for five fuel cycles and got the average fuel burnup or varies as 50MW-day/kgU, a conclusion is made that the WWER-440 fuel burnup can be increased at least to average burnups of 55-58 MW-day/kgU per fuel assembly (Authors)

TRIGA criticality experiment for testing burn-up calculations

International Nuclear Information System (INIS)

Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz

1999-01-01

A criticality experiment with partly burned TRIGA fuel is described. 20 wt % enriched standard TRIGA fuel elements initially containing 12 wt % U are used. Their average burn-up is 1.4 MWd. Fuel element burn-up is calculated in 2-D four group diffusion approximation using TRIGLAV code. The burn-up of several fuel elements is also measured by reactivity method. The excess reactivity of several critical and subcritical core configurations is measured. Two core configurations contain the same fuel elements in the same arrangement as were used in the fresh TRIGA fuel criticality experiment performed in 1991. The results of the experiment may be applied for testing the computer codes used for fuel burn-up calculations. (author)

Optimum Discharge Burnup and Cycle Length for PWRs

International Nuclear Information System (INIS)

Secker, Jeffrey R.; Johansen, Baard J.; Stucker, David L.; Ozer, Odelli; Ivanov, Kostadin; Yilmaz, Serkan; Young, E.H.

2005-01-01

This paper discusses the results of a pressurized water reactor fuel management study determining the optimum discharge burnup and cycle length. A comprehensive study was performed considering 12-, 18-, and 24-month fuel cycles over a wide range of discharge burnups. A neutronic study was performed followed by an economic evaluation. The first phase of the study limited the fuel enrichments used in the study to 235 U consistent with constraints today. The second phase extended the range of discharge burnups for 18-month cycles by using fuel enriched in excess of 5 wt%. The neutronic study used state-of-the-art reactor physics methods to accurately determine enrichment requirements. Energy requirements were consistent with today's high capacity factors (>98%) and short (15-day) refueling outages. The economic evaluation method considers various component costs including uranium, conversion, enrichment, fabrication and spent-fuel storage costs as well as the effect of discounting of the revenue stream. The resulting fuel cycle costs as a function of cycle length and discharge burnup are presented and discussed. Fuel costs decline with increasing discharge burnup for all cycle lengths up to the maximum discharge burnup considered. The choice of optimum cycle length depends on assumptions for outage costs

Development of high burnup nuclear fuel technology

International Nuclear Information System (INIS)

Suk, Ho Chun; Kang, Young Hwan; Jung, Jin Gone; Hwang, Won; Park, Zoo Hwan; Ryu, Woo Seog; Kim, Bong Goo; Kim, Il Gone

1987-04-01

The objectives of the project are mainly to develope both design and manufacturing technologies for 600 MWe-CANDU-PHWR-type high burnup nuclear fuel, and secondly to build up the foundation of PWR high burnup nuclear fuel technology on the basis of KAERI technology localized upon the standard 600 MWe-CANDU- PHWR nuclear fuel. So, as in the first stage, the goal of the program in the last one year was set up mainly to establish the concept of the nuclear fuel pellet design and manufacturing. The economic incentives for high burnup nuclear fuel technology development are improvement of fuel utilization, backend costs plant operation, etc. Forming the most important incentives of fuel cycle costs reduction and improvement of power operation, etc., the development of high burnup nuclear fuel technology and also the research on the incore fuel management and safety and technologies are necessary in this country

Fundamental burn-up mode in a pebble-bed type reactor

International Nuclear Information System (INIS)

Chen, Xue-Nong; Kiefhaber, Edgar; Maschek, Werner

2008-01-01

This paper deals with a pebble-bed type reactor, in which the fuel is loaded from one side (top) and discharged from the other side (bottom). A boundary value problem of a single group diffusion equation coupled with simplified burn-up equations is studied, where the natural radioactive decay processes are neglected in the burn-up modelling. An asymptotic burning wave solution is found analytically in the one-dimensional case, which is called as fundamental burn-up mode. Among this solution family there are two particular cases, namely, a classic fundamental solution with a zero burn-up and a partial solitary burn-up wave solution with a highest burn-up. An example of Th-U conversion is considered and the solutions are presented in order to show the mechanism of the burning wave. (author)

Benefits of actinide-only burnup credit for shutdown PWRs

International Nuclear Information System (INIS)

Lancaster, D.; Fuentes, E.; Kang, C.; Rivard, D.

1998-02-01

Owners of PWRs that are shutdown prior to resolution of interim storage or permanent disposal issues have to make difficult decisions on what to do with their spent fuel. Maine Yankee is currently evaluating multiple options for spent fuel storage. Their spent fuel pool has 1,434 assemblies. In order to evaluate the value to a utility of actinide-only burnup credit, analysis of the number of canisters required with and without burnup credit was made. In order to perform the analysis, loading curves were developed for the Holtec Hi-Star 100/MPC-32. The MPC-32 is hoped to be representative of future burnup credit designs from many vendors. The loading curves were generated using the actinide-only burnup credit currently under NRC review. The canister was analyzed for full loading (32 assemblies) and with partial loadings of 30 and 28 assemblies. If no burnup credit is used the maximum capacity was assumed to be 24 assemblies. this reduced capacity is due to the space required for flux traps which are needed to sufficiently reduce the canister reactivity for the fresh fuel assumption. Without burnup credit the 1,343 assemblies would require 60 canisters. If all the fuel could be loaded into the 32 assembly canisters only 45 canisters would be required. Although the actinide-only burnup credit approach is very conservative, the total number of canisters required is only 47 which is only two short of the minimum possible number of canisters. The utility is expected to buy the canister and the storage overpack. A reasonable cost estimate for the canister plus overpack is $500,000. Actinide-only burnup credit would save 13 canisters and overpacks which is a savings of about $6.5 million. This savings is somewhat reduced since burnup credit requires a verification measurement of burnup. The measurement costs for these assemblies can be estimated as about $1 million. The net savings would be $5.5 million

SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

International Nuclear Information System (INIS)

BSC

2004-01-01

Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier

Alloy development for high burnup cladding (PWR)

Energy Technology Data Exchange (ETDEWEB)

Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-04-01

An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

Preparation of data relevant to ''Equivalent Uniform Burnup'' and Equivalent Initial Enrichment'' for burnup credit evaluation

Energy Technology Data Exchange (ETDEWEB)

Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan)

2001-11-01

Based on the PWR spent fuel composition data measured at JAERI, two kinds of simplified methods such as ''Equivalent Uniform Burnup'' and ''Equivalent Initial Enrichment'' have been introduced. And relevant evaluation curves have been prepared for criticality safety evaluation of spent fuel storage pool and transport casks, taking burnup of spent fuel into consideration. These simplified methods can be used to obtain an effective neutron multiplication factor for a spent fuel storage/transportation system by using the ORIGEN2.1 burnup code and the KENO-Va criticality code without considering axial burnup profile in spent fuel and other various factors introducing calculated errors. ''Equivalent Uniform Burnup'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis, in which the experimentally obtained isotopic composition together with a typical axial burnup profile and various factors such as irradiation history are considered on the conservative side. On the other hand, Equivalent Initial Enrichment'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis such as above when it is used in the so called fresh fuel assumption. (author)

Light a CANDLE. An innovative burnup strategy of nuclear reactors

International Nuclear Information System (INIS)

Sekimoto, Hiroshi

2005-11-01

CANDLE is a new burnup strategy for nuclear reactors, which stands for Constant Axial Shape of Neutron Flux, Nuclide Densities and Power Shape During Life of Energy Production. When this candle-like burnup strategy is adopted, although the fuel is fixed in a reactor core, the burning region moves, at a speed proportionate to the power output, along the direction of the core axis without changing the spatial distribution of the number density of the nuclides, neutron flux, and power density. Excess reactivity is not necessary for burnup and the shape of the power distribution and core characteristics do not change with the progress of burnup. It is not necessary to use control rods for the control of the burnup. This booklet described the concept of the CANDLE burnup strategy with basic explanations of excess neutrons and its specific application to a high-temperature gas-cooled reactor and a fast reactor with excellent neutron economy. Supplementary issues concerning the initial core and high burnup were also referred. (T. Tanaka)

Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

International Nuclear Information System (INIS)

Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi; Kaneko, Toshiyuki.

1996-12-01

The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code 'MULTI-KENO' and the routine for the burnup calculation of the one dimensional burnup code 'UNITBURN'. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

Energy Technology Data Exchange (ETDEWEB)

Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Toshiyuki

1996-12-01

The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code `MULTI-KENO` and the routine for the burnup calculation of the one dimensional burnup code `UNITBURN`. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

Burnup calculation in microcells of high conversion reactors

International Nuclear Information System (INIS)

Gomez, S.E.; Salvatore, M.; Patino, N.E.; Abbate, M.J.

1991-01-01

The development of high converter reactors (HCR) requires careful burnup calculations because their main goals are reach high discharge burnup levels (Up to 50 GWd/T) and a close to one conversion ratio. Then, it is necessary a revision of design elements used for this type of calculation. In this work, a burnup module (BUM) developed in order to use nuclear data directly from evaluated data files is presented; these was included in the AMPX system. (author)

A PWR Thorium Pin Cell Burnup Benchmark

Energy Technology Data Exchange (ETDEWEB)

Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

2000-05-01

As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

Burnup analysis of the power reactor, 2

International Nuclear Information System (INIS)

Ezure, Hideo

1975-09-01

In burnup analysis of JPDR-1 with FLARE, it was found to have problems. The program FLORA was developed for solution of the problems. By their bench mark tests FLORA was found to be useful for three-dimensional thermal-hydro-dynamic analysis of BWRs. It was applied to analysis of the burnup of JPDR-1. The input data and option of FLORA were corrected on referring to the results of gammer probe tests for JPDR-1. The void, source and burnup distributions were calculated each month during the operation. The burnup distribution in three assemblies revealed by a destructive test agrees better with that by FLORA than by FLARE. It was shown that the distortion of power distribution around the control rods by FLORA was smaller and closer to that by the gammer probe tests than by FLARE, and the connector of fuel assemblies and the plugs in the reflector had much influence on the power distribution. (auth.)

Automated generation of burnup chain for reactor analysis applications

International Nuclear Information System (INIS)

Tran Viet Phu; Tran Hoai Nam; Akio Yamamoto; Tomohiro Endo

2015-01-01

This paper presents the development of an automated generation of a new burnup chain for reactor analysis applications. The JENDL FP Decay Data File 2011 and Fission Yields Data File 2011 were used as the data sources. The nuclides in the new chain are determined by restrictions of the half-life and cumulative yield of fission products or from a given list. Then, decay modes, branching ratios and fission yields are recalculated taking into account intermediate reactions. The new burnup chain is output according to the format for the SRAC code system. Verification was performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Further development and applications are being planned with the burnup chain code. (author)

A Monte Carlo burnup code linking MCNP and REBUS

International Nuclear Information System (INIS)

Hanan, N.A.; Olson, A.P.; Pond, R.B.; Matos, J.E.

1998-01-01

The REBUS-3 burnup code, used in the anl RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented. (author)

Burnup credit applications in a high-capacity truck cask

International Nuclear Information System (INIS)

Boshoven, J.K.

1992-09-01

General Atomics (GA) has designed two legal weight truck (LWT) casks, the GA-4 and GA-9, to carry four pressurized-water-reactor (PWR) and nine boiling-water-reactor (BWR) fuel assemblies, respectively. GA plans to submit applications for certification to the US Nuclear Regulatory Commission (NRC) for the two casks in mid-1993. GA will include burnup credit analysis in the Safety Analysis Report for Packaging (SARP) for the GA-4 Cask. By including burnup credit in the criticality safety analysis for PWR fuels with initial enrichments above 3% U-235, public and occupation risks are reduced and cost savings are realized. The GA approach to burnup credit analysis incorporates the information produced in the US Department of Energy Burnup Credit Program. This paper describes the application of burnup credit to the criticality control design of the GA-4 Cask

EVOLUT - a computer program for fast burnup evaluation

International Nuclear Information System (INIS)

Craciunescu, T.; Dobrin, R.; Stamatescu, L.; Alexa, A.

1999-01-01

EVOLUT is a computer program for burnup evaluation. The input data consist on the one hand of axial and radial gamma-scanning profiles (for the experimental evaluation of the number of nuclei of a fission product - the burnup monitor - at the end of irradiation) and on the other hand of the history of irradiation (the time length and values proportional to the neutron flux for each step of irradiation). Using the equation of evolution of the burnup monitor the flux values are iteratively adjusted, by a multiplier factor, until the calculated number of nuclei is equal to the experimental one. The flux values are used in the equation of evolution of the fissile and fertile nuclei to determine the fission number and consequently the burnup. EVOLUT was successfully used in the analysis of several hundreds of CANDU and TRIGA-type fuel rods. We appreciate that EVOLUT is a useful tool in the burnup evaluation based on gamma spectrometry measurements. EVOLUT can be used on an usual AT computer and in this case the results are obtained in a few minutes. It has an original and user-friendly graphical interface and it provides also output in script MATLAB files for graphical representation and further numerical analysis. The computer program needs simple data and it is valuable especially when a large number of burnup analyses are required quickly. (authors)

Sophistication of burnup analysis system for fast reactor

International Nuclear Information System (INIS)

Yokoyama, Kenji; Hirai, Yasushi; Hyoudou, Hideaki; Tatsumi, Masahiro

2010-02-01

Improvement on prediction accuracy for neutronics property of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified constants library as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores, however, improvement of not only static properties but also burnup properties is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup properties using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous research, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for production systems. In the present study, we implemented functions for cell calculations and burnup calculations. With this, whole steps in analysis can be carried out with only this system. In addition, we modified the specification of user input to improve the convenience of this system. Since implementations being done so

Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up

DEFF Research Database (Denmark)

Carlsen, H.

1980-01-01

Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...

Fuel analysis code FAIR and its high burnup modelling capabilities

International Nuclear Information System (INIS)

Prasad, P.S.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

1995-01-01

A computer code FAIR has been developed for analysing performance of water cooled reactor fuel pins. It is capable of analysing high burnup fuels. This code has recently been used for analysing ten high burnup fuel rods irradiated at Halden reactor. In the present paper, the code FAIR and its various high burnup models are described. The performance of code FAIR in analysing high burnup fuels and its other applications are highlighted. (author). 21 refs., 12 figs

Appropriate burnup measurements for transportation burnup credit

International Nuclear Information System (INIS)

Lancaster, D.; Fuentes, E.

1997-01-01

This paper addresses two of the measurement specifications used in analyzing spent fuel packages to gain burnup credit. The philosophy and calculation of rejection criteria and measurement accuracy are discussed. Any assembly for which the declared measured value and reactor record value deviate by more than 10% will be rejected. Measurement accuracy requirements are established for dependent and independent systems. The requirements have been tested and are achievable, ensuring safe operation without extra cost. 6 refs

Burnup verification using the FORK measurement system

International Nuclear Information System (INIS)

Ewing, R.I.

1994-01-01

Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. The FORK measurement system, designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program, has been used to verify reactor site records for burnup and cooling time for many years. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. This report deals with the application of the FORK system to burnup credit operations based on measurements performed on spent fuel assemblies at the Oconee Nuclear Station of Duke Power Company

MTR fuel element burn-up measurements by the reactivity method

International Nuclear Information System (INIS)

Zuniga, A.; Cuya, T.R.; Ravnik, M.

2003-01-01

Fuel element burn-up was measured by the reactivity method in the 10 MW Peruvian MTR reactor RP-10. The main purpose of the experiment was testing the reactivity method for an MTR reactor as the reactivity method was originally developed for TRIGA reactors. The reactivity worth of each measured fuel element was measured in its original core position in order to measure the burn-up of the fuel elements that were part of the experimental core. The burn-up of each measured fuel element was derived by interpolating its reactivity worth from the reactivity worth of two reference fuel elements of known burn-up, whose reactivity worth was measured in the position of the measured fuel element. The accuracy of the method was improved by separating the reactivity effect of burn-up from the effect of the position in the core. The results of the experiment showed that the modified reactivity method for fuel element burn-up determination could be applied also to MTR reactors. (orig.)

Fuel burnup analysis for the Moroccan TRIGA research reactor

International Nuclear Information System (INIS)

El Bakkari, B.; El Bardouni, T.; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Zoubair, M.

2013-01-01

Highlights: ► A fuel burnup analysis of the 2 MW TRIGA MARK II Moroccan research reactor was established. ► Burnup calculations were done by means of the in-house developed burnup code BUCAL1. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► The reactor life time was found to be 3360 MW h considering full power operating conditions. ► Power factors and fluxes of the in-core irradiation positions are strongly affected by burnup. -- Abstract: The fundamental advantage and main reason to use Monte Carlo methods for burnup calculations is the possibility to generate extremely accurate burnup dependent one group cross-sections and neutron fluxes for arbitrary core and fuel geometries. Yet, a set of values determined for a material at a given position and time remains accurate only in a local region, in which neutron spectrum and flux vary weakly — and only for a limited period of time, during which changes of the local isotopic composition are minor. This paper presents the approach of fuel burnup evaluation used at the Moroccan TRIGA MARK II research reactor. The approach is essentially based upon the utilization of BUCAL1, an in-house developed burnup code. BUCAL1 is a FORTRAN computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in nuclear reactors. The code was developed to incorporate the neutron absorption reaction tally information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The fuel cycle length and changes in several core parameters such as: core excess reactivity, control rods position, fluxes at the irradiation positions, axial and radial power factors and other parameters are estimated. Besides, this study gives valuable insight into the behavior of the reactor and will ensure better utilization and operation of the reactor during its life-time and it will allow the establishment of

Analysis of the behavior under irradiation of high burnup nuclear fuels with the computer programs FRAPCON and FRAPTRAN

Energy Technology Data Exchange (ETDEWEB)

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

2017-11-01

The objective of this paper is to verify the validity and accuracy of the results provided by computer programs FRAPCON-3.4a and FRAPTRAN-1.4, used in the simulation process of the irradiation behavior of Pressurized Water Reactors (PWR) fuel rods, in steady-state and transient operational conditions at high burnup. To achieve this goal, the results provided by these computer simulations are compared with experimental data available in the database FUMEX III. Through the results, it was found that the computer programs used have a good ability to predict the operational behavior of PWR fuel rods in high burnup steady-state conditions and under the influence of Reactivity Initiated Accident (RIA). (author)

The Width of High Burnup Structure in LWR UO2 Fuel

International Nuclear Information System (INIS)

Koo, Yang-Hyun; Lee, Byung-Ho; Oh, Jae-Yong; Sohn, Dong-Seong

2007-01-01

The measured data available in the open literature on the width of high burnup structure (HBS) in LWR UO 2 fuel were analyzed in terms of pellet average burnup, enrichment, and grain size. Dependence of the HBS width on pellet average burnup was shown to be divided into three regions; while the HBS width is governed by accumulation of fission damage (i.e., burnup) for burnup below 60 GWd/tU, it seems to be restricted to some limiting value of around 1.5 mm for burnup above 75 GWd/tU due to high temperature which might have caused extensive annealing of irradiation damage. As for intermediate burnup between 60 and 75 GWd/tU, although temperature would not have been so high as to induce extensive annealing, the microstructural damage could have been partly annealed, resulting in the reduction of the HBS width. It was found that both enrichment and grain size also affects the HBS width. However, as long as the pellet average burnup is lower than about 75 GWd/tU, the effect does not appear to be significant for the enrichment and grain size that are typically used in current LWR fuel. (authors)

Thermal transport characterization of hexagonal boron nitride nanoribbons using molecular dynamics simulation

Directory of Open Access Journals (Sweden)

Asir Intisar Khan

2017-10-01

Full Text Available Due to similar atomic bonding and electronic structure to graphene, hexagonal boron nitride (h-BN has broad application prospects such as the design of next generation energy efficient nano-electronic devices. Practical design and efficient performance of these devices based on h-BN nanostructures would require proper thermal characterization of h-BN nanostructures. Hence, in this study we have performed equilibrium molecular dynamics (EMD simulation using an optimized Tersoff-type interatomic potential to model the thermal transport of nanometer sized zigzag hexagonal boron nitride nanoribbons (h-BNNRs. We have investigated the thermal conductivity of h-BNNRs as a function of temperature, length and width. Thermal conductivity of h-BNNRs shows strong temperature dependence. With increasing width, thermal conductivity increases while an opposite pattern is observed with the increase in length. Our study on h-BNNRs shows considerably lower thermal conductivity compared to GNRs. To elucidate these aspects, we have calculated phonon density of states for both h-BNNRs and GNRs. Moreover, using EMD we have explored the impact of different vacancies, namely, point vacancy, edge vacancy and bi-vacancy on the thermal conductivity of h-BNNRs. With varying percentages of vacancies, significant reduction in thermal conductivity is observed and it is found that, edge and point vacancies are comparatively more destructive than bi-vacancies. Such study would contribute further into the growing interest for accurate thermal transport characterization of low dimensional nanostructures.

Nitride alloy layer formation of duplex stainless steel using nitriding process

Science.gov (United States)

Maleque, M. A.; Lailatul, P. H.; Fathaen, A. A.; Norinsan, K.; Haider, J.

2018-01-01

Duplex stainless steel (DSS) shows a good corrosion resistance as well as the mechanical properties. However, DSS performance decrease as it works under aggressive environment and at high temperature. At the mentioned environment, the DSS become susceptible to wear failure. Surface modification is the favourable technique to widen the application of duplex stainless steel and improve the wear resistance and its hardness properties. Therefore, the main aim of this work is to nitride alloy layer on the surface of duplex stainless steel by the nitriding process temperature of 400°C and 450°C at different time and ammonia composition using a horizontal tube furnace. The scanning electron microscopy and x-ray diffraction analyzer are used to analyse the morphology, composition and the nitrided alloy layer for treated DSS. The micro hardnesss Vickers tester was used to measure hardness on cross-sectional area of nitrided DSS. After nitriding, it was observed that the hardness performance increased until 1100 Hv0.5kgf compared to substrate material of 250 Hv0.5kgf. The thickness layer of nitride alloy also increased from 5μm until 100μm due to diffusion of nitrogen on the surface of DSS. The x-ray diffraction results showed that the nitride layer consists of iron nitride, expanded austenite and chromium nitride. It can be concluded that nitride alloy layer can be produced via nitriding process using tube furnace with significant improvement of microstructural and hardness properties.

Determination of axial profit performed burnup credit by SCALE 4.3-system

International Nuclear Information System (INIS)

Miro, R.; Verdu, G.; Munoz-Cobo, J. L.

1998-01-01

SCALE 4.3 is a modular code system designed for realizing standard computational analysis for licensing evaluation. Since now, spent fuel storage pools criticality analysis have been done considering this fuel as fresh, with its maximum enrichment. With burnup credit we can obtain cheaper and compact configurations. The procedure for calculating a spent fuel storage consists of a burnup calculation plus a criticality calculation. We can perform a conservative approximation for the burnup calculations using 1-D results, but, besides the geometry configurations for the 3-D criticality calculation. we need an appropriate approximation to model the burnup axial variation. We assume that for a burnup profile set, the most conservative profile is between the lower and the upper range of this profile, set. We consider only combinations of the maximum and minimum burnup in each axial region, for each burnup range. This gives an estimation of the different burnup shapes effect and the general characteristics of the most conservative shape. (Author) 6 refs

Nuclear fuel burn-up economy

International Nuclear Information System (INIS)

Matausek, M.

1984-01-01

In the period 1981-1985, for the needs of Utility Organization, Beograd, and with the support of the Scientific Council of SR Srbija, work has been performed on the study entitled 'Nuclear Fuel Burn-up Economy'. The forst [phase, completed during the year 1983 comprised: comparative analysis of commercial NPP from the standpoint of nuclear fuel requirements; development of methods for fuel burn-up analysis; specification of elements concerning the nuclear fuel for the tender documentation. The present paper gives the short description of the purpose, content and results achieved in the up-to-now work on the study. (author)

Systemization of burnup sensitivity analysis code (2) (Contract research)

International Nuclear Information System (INIS)

Tatsumi, Masahiro; Hyoudou, Hideaki

2008-08-01

Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion

Comparative sodium void effects for different advanced liquid metal reactor fuel and core designs

International Nuclear Information System (INIS)

Dobbin, K.D.; Kessler, S.F.; Nelson, J.V.; Gedeon, S.R.; Omberg, R.P.

1991-01-01

An analysis of metal-, oxide-, and nitride-fueled advanced liquid metal reactor cores was performed to investigate the calculated differences in sodium void reactivity, and to determine the relationship between sodium void reactivity and burnup reactivity swing using the three fuel types. The results of this analysis indicate that nitride fuel has the least positive sodium void reactivity for any given burnup reactivity swing. Thus, it appears that a good design compromise between transient overpower and loss of flow response is obtained using nitride fuel. Additional studies were made to understand these and other nitride advantages. (author)

Burnup calculation code system COMRAD96

International Nuclear Information System (INIS)

Suyama, Kenya; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu.

1997-06-01

COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)

Heterostructures for Increased Quantum Efficiency in Nitride LEDs

Energy Technology Data Exchange (ETDEWEB)

Davis, Robert F. [Carnegie Mellon Univ., Pittsburgh, PA (United States)

2010-09-30

Task 1. Development of an advanced LED simulator useful for the design of efficient nitride-based devices. Simulator will contain graphical interface software that can be used to specify the device structure, the material parameters, the operating conditions and the desired output results. Task 2. Theoretical and experimental investigations regarding the influence on the microstructure, defect concentration, mechanical stress and strain and IQE of controlled changes in the chemistry and process route of deposition of the buffer layer underlying the active region of nitride-based blue- and greenemitting LEDs. Task 3. Theoretical and experimental investigations regarding the influence on the physical properties including polarization and IQE of controlled changes in the geometry, chemistry, defect density, and microstructure of components in the active region of nitride-based blue- and green-emitting LEDs. Task 4. Theoretical and experimental investigations regarding the influence on IQE of novel heterostructure designs to funnel carriers into the active region for enhanced recombination efficiency and elimination of diffusion beyond this region. Task 5. Theoretical and experimental investigations regarding the influence of enhanced p-type doping on the chemical, electrical, and microstructural characteristics of the acceptor-doped layers, the hole injection levels at Ohmic contacts, the specific contact resistivity and the IQE of nitride-based blue- and green-emitting LEDs. Development and optical and electrical characterization of reflective Ohmic contacts to n- and p-type GaN films.

Determination of burn-up of irradiated nuclear fuels using mass spectrometry

International Nuclear Information System (INIS)

Jagadish Kumar, S.; Telmore, V.M.; Shah, R.V.; Sasi Bhushan, K.; Paul, Sumana; Kumar, Pranaw; Rao, Radhika M.; Jaison, P.G.

2017-01-01

Burn-up defined as the atom percent fission, is a vital parameter used for assessing the performance of nuclear fuel during its irradiation in the reactor. Accurate data on the actinide isotopes are also essential for the reliable accountability of nuclear materials and for nuclear safeguards. Both destructive and non-destructive methods are employed in the post-irradiation analysis for the burn-up measurements. Though non-destructive methods are preferred from the point view of remote handling of irradiated fuels with high radioactivity, they do not provide the high accuracy as achieved by the chemical analysis methods. Thus destructive radiochemical and chemical analyses are still the established reference methods for accurate and reliable burn-up determination of irradiated nuclear fuels. In the destructive method, burn-up of irradiated nuclear fuel is determined by correlating the amount of a fission product formed during irradiation with that of heavy elements. Thus the destructive experimental determination of burn-up involves the dissolution of irradiated fuel samples followed by the separation and determination of heavy elements and fission product(s) to be used as burn-up monitor(s). Another approach for the experimental determination of burn-up is based on the changes in the abundances of the heavy element isotopes. A widely accepted method for burn-up determination is based on stable "1"4"8Nd and "1"3"9La as burn-up monitors. Several properties such as non-volatility, nearly same yields for thermal fissions of "2"3"5U and "2"3"9Pu etc justifies the selection of "1"4"8Nd as a burn-up monitor

Structural properties of iron nitride on Cu(100): An ab-initio molecular dynamics study

KAUST Repository

Heryadi, Dodi

2011-01-01

Due to their potential applications in magnetic storage devices, iron nitrides have been a subject of numerous experimental and theoretical investigations. Thin films of iron nitride have been successfully grown on different substrates. To study the structural properties of a single monolayer film of FeN we have performed an ab-initio molecular dynamics simulation of its formation on a Cu(100) substrate. The iron nitride layer formed in our simulation shows a p4gm(2x2) reconstructed surface, in agreement with experimental results. In addition to its structural properties, we are also able to determine the magnetization of this thin film. Our results show that one monolayer of iron nitride on Cu(100) is ferromagnetic with a magnetic moment of 1.67 μ B. © 2011 Materials Research Society.

Modelling the high burnup UO2 structure in LWR fuel

International Nuclear Information System (INIS)

Lassmann, K.; Walker, C.T.; Laar, J. van de; Lindstroem, F.

1995-01-01

The concept of a burnup threshold for the formation of the high burnup UO 2 structure (HBS) is supported by experimental data, which also reveal that a transition zone exists between the normal UO 2 structure and the fully developed HBS. From the analysis of radial xenon profiles measured by EPMA a threshold burnup is obtained in the range 60-75 GW d/t U. The lower value is considered to be the threshold for the onset of the HBS and the higher value the threshold for the fully developed HBS. Xenon depletion in the transition zone and the fully developed HBS can be described by a simple model. At local burnups above 120 GW d/t U the xenon generated is in equilibrium with the xenon lost to the fission gas pores and the concentration does not fall below 0.25 wt%. The TRANSURANUS burnup model TUBRNP predicts reasonably well the penetration of the HBS and the associated xenon depletion up to a cross section average burnup of approximately 70 GW d/t U. (orig.)

An economic evaluation of a storage system for casks with burnup credit

International Nuclear Information System (INIS)

Mimura, Masahiro; Tsuda, Kazuaki; Yamada, Nobuyuki; O-iwa, Akio.

1993-01-01

It is generally recognized that casks designed with burnup credit are more economical than those without burnup credit. To estimate how much more economical they are, we made conceptual designs of transport/storage casks with and without burnup credit for PWR and BWR fuels of various uranium enrichment. The casks were designed to contain the maximum number of fuel assemblies under the necessary weight and dimensional limitations as well as the criticality and shielding criteria. The results showed that approximately 8 % to 44 % more fuel assemblies could be contained in casks with burnup credit. We then evaluated the economy of cask storage system incorporating the cask designs obtained above both with and without burnup credit. The results showed that the cost of storing casks with burnup credit is approximately 7 % to 30 % less expensive than storing casks without burnup credit. (J.P.N.)

Status of burnup credit implementation in Switzerland

International Nuclear Information System (INIS)

Grimm, P.

1998-01-01

Burnup credit is currently not used for the storage of spent fuel in the reactor pools in Switzerland, but credit is taken for integral burnable absorbers. Interest exists to take credit of burnup in future for the storage in a central away-from-reactor facility presently under construction. For spent fuel transports to foreign reprocessing plants the regulations of the receiving countries must be applied in addition to the Swiss licensing criteria. Burnup credit has been applied by one Swiss PWR utility for such transports in a consistent manner with the licensing practice in the receiving countries. Measurements of reactivity worths of small spent fuel samples in a Swiss zero-power research reactor are at an early stage of planning. (author)

Two dimensional burn-up calculation of TRIGA core

International Nuclear Information System (INIS)

Persic, A.; Ravnik, M.; Slavic, S.

1996-01-01

TRIGLAV is a new computer program for burn-up calculation of mixed core of research reactors. The code is based on diffusion model in two dimensions and iterative procedure is applied for its solution. The material data used in the model are calculated with the transport program WIMS. In regard to fission density distribution and energy produced by the reactor the burn-up increment of fuel elements is determined. In this paper the calculation model of diffusion constants and burn-up calculation are described and some results of calculations for TRIGA MARK II reactor are presented. (author)

Impact of extended burnup on the nuclear fuel cycle

International Nuclear Information System (INIS)

1993-04-01

The Advisory Group Meeting was held in Vienna from 2 to 5 December 1991, to review, analyse, and discuss the effects of burnup extension in both light and heavy water reactors on all aspects of the fuel cycle. Twenty experts from thirteen countries participated in this meeting. There was consensus that both economic and environmental benefits are driving forces toward the achievement of higher burnups and that the present trend of burnup extension may be expected to continue. The extended burnup has been considered for the three main stages of the fuel cycle: the front end, in-reactor issues and the back end. Thirteen papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

Burnup calculation code system COMRAD96

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu

1997-06-01

COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, `Cross Section Treatment`, `Generation and Depletion Calculation`, and `Post Process`. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the {gamma} Spectrum on a terminal. This report is the general description and user`s manual of COMRAD96. (author)

Analyzing the BWR rod drop accident in high-burnup cores

International Nuclear Information System (INIS)

Diamond, D.J.; Neymotin, L.; Kohut, P.

1995-01-01

This study was undertaken for the US Nuclear Regulatory Commission to determine the fuel enthalpy during a rod drop accident (RDA) for cores with high burnup fuel. The calculations were done with the RAMONA-4B code which models the core with 3-dimensional neutron kinetics and multiple parallel coolant channels. The calculations were done with a model for a BWR/4 with fuel bundles having burnups up to 30 GWd/t and also with a model with bundle burnups to 60 GWd/t. This paper also discusses potential sources of uncertainty in calculations with high burnup fuel. One source is the ''rim'' effect which is the extra large peaking of the power distribution at the surface of the pellet. This increases the uncertainty in reactor physics and heat conduction models that assume that the energy deposition has a less peaked spatial distribution. Two other sources of uncertainty are the result of the delayed neutron fraction decreasing with burnup and the positive moderator temperature feedback increasing with burnup. Since these effects tend to increase the severity of the event, an RDA calculation for high burnup fuel will underpredict the fuel enthalpy if the effects are not properly taken into account. Other sources of uncertainty that are important come from the initial conditions chosen for the RDA. This includes the initial control rod pattern as well as the initial thermal-hydraulic conditions

Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

Energy Technology Data Exchange (ETDEWEB)

Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

2016-10-15

Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

International Nuclear Information System (INIS)

Barkauskas, V.; Plukiene, R.; Plukis, A.

2016-01-01

Highlights: • RBMK-1500 fuel burn-up impact on k_e_f_f in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k_e_f_f in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k_e_f_f) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality safety.

Development of burnup methods and capabilities in Monte Carlo code RMC

International Nuclear Information System (INIS)

She, Ding; Liu, Yuxuan; Wang, Kan; Yu, Ganglin; Forget, Benoit; Romano, Paul K.; Smith, Kord

2013-01-01

Highlights: ► The RMC code has been developed aiming at large-scale burnup calculations. ► Matrix exponential methods are employed to solve the depletion equations. ► The Energy-Bin method reduces the time expense of treating ACE libraries. ► The Cell-Mapping method is efficient to handle massive amounts of tally cells. ► Parallelized depletion is necessary for massive amounts of burnup regions. -- Abstract: The Monte Carlo burnup calculation has always been a challenging problem because of its large time consumption when applied to full-scale assembly or core calculations, and thus its application in routine analysis is limited. Most existing MC burnup codes are usually external wrappers between a MC code, e.g. MCNP, and a depletion code, e.g. ORIGEN. The code RMC is a newly developed MC code with an embedded depletion module aimed at performing burnup calculations of large-scale problems with high efficiency. Several measures have been taken to strengthen the burnup capabilities of RMC. Firstly, an accurate and efficient depletion module called DEPTH has been developed and built in, which employs the rational approximation and polynomial approximation methods. Secondly, the Energy-Bin method and the Cell-Mapping method are implemented to speed up the transport calculations with large numbers of nuclides and tally cells. Thirdly, the batch tally method and the parallelized depletion module have been utilized to better handle cases with massive amounts of burnup regions in parallel calculations. Burnup cases including a PWR pin and a 5 × 5 assembly group are calculated, thereby demonstrating the burnup capabilities of the RMC code. In addition, the computational time and memory requirements of RMC are compared with other MC burnup codes.

Optimization of TRU burnup in modular helium reactor

International Nuclear Information System (INIS)

Yonghee, Kim; Venneri, F.

2007-01-01

An optimization study of a single-pass TRU (transuranic) deep-burn (DB) has been performed for a block-type MHR (Modular Helium Reactor) proposed by General Atomics. Assuming a future equilibrium scenario of advanced LWRs, a high-burnup TRU vector is considered: 50 GWD/MTU and 5-year cooling. For 3-D equilibrium cores, the performance analysis is done by using a continuous energy Monte Carlo depletion code MCCARD. The core optimization is performed from the viewpoints of the core configuration, fuel management, TRISO fuel specification, and neutron spectrum. With regard to core configuration, two annular cores are investigated in terms of the neutron economy. A conventional radial shuffling scheme of fuel blocks is compared with an axial block shuffling strategy in terms of the fuel burnup and core power distributions. The impact of the kernel size of TRISO fuel is evaluated and a diluted kernel, instead of a conventional concentrated kernel, is introduced to maximize the TRU burnup by reducing the self-shielding effects of TRISO fuels. A higher graphite density is evaluated in terms of the fuel burnup. In addition, it is shown that the core power distribution can be effectively controlled by zoning of the packing fraction of TRISO fuels. We also have shown that a long-cycle DB-MHR core can be designed by using a small batch size for fuel reloading, at the expense of a marginal decrease of the TRU discharge burnup. Depending on the fuel management scheme, fuel specifications, and core parameters, the TRU burnup in an optimized DB-MHR core is over 60% in a single-pass irradiation campaign. (authors)

Parametric neutronic analyses related to burnup credit cask design

International Nuclear Information System (INIS)

Parks, C.V.

1989-01-01

The consideration of spent fuel histories (burnup credit) in the design of spent fuel shipping casks will result in cost savings and public risk benefits in the overall fuel transportation system. The purpose of this paper is to describe the depletion and criticality analyses performed in conjunction with and supplemental to the referenced analysis. Specifically, the objectives are to indicate trends in spent fuel isotopic composition with burnup and decay time; provide spent fuel pin lattice values as a function of burnup, decay time, and initial enrichment; demonstrate the variation of k eff for infinite arrays of spent fuel assemblies separated by generic cask basket designs (borated and unborated) of varying thicknesses; and verify the potential cask reactivity margin available with burnup credit via analysis with generic cask models

WEAR PERFORMANCE OPTIMIZATION OF SILICON NITRIDE USING GENETIC AND SIMULATED ANNEALING ALGORITHM

Directory of Open Access Journals (Sweden)

SACHIN GHALME

2017-12-01

Full Text Available Replacing damaged joint with the suitable alternative material is a prime requirement in a patient who has arthritis. Generation of wear particles in the artificial joint during action or movement is a serious issue and leads to aseptic loosening of joint. Research in the field of bio-tribology is trying to evaluate materials with minimum wear volume loss so as to extend joint life. Silicon nitride (Si3N4 is non-oxide ceramic suggested as a new alternative for hip/knee joint replacement. Hexagonal Boron Nitride (hBN is recommended as a solid additive lubricant to improve the wear performance of Si3N4 . In this paper, an attempt has been made to evaluate the optimum combination of load and % volume of hBN in Si3N4 to minimize wear volume loss (WVL. The experiments were conducted according to Design of Experiments (DoE – Taguchi method and a mathematical model is developed. Further, this model is processed with Genetic Algorithm (GA and Simulated Annealing (SA to find out the optimum percentage of hBN in Si3N4 to minimize wear volume loss against Alumina (Al2O3 counterface. Taguchi method presents 15 N load and 8% volume of hBN to minimize WVL of Si3N4 . While GA and SA optimization offer 11.08 N load, 12.115% volume of hBN and 11.0789 N load, 12.128% volume of hBN respectively to minimize WVL in Si3N4. .

Oxygen stoichiometry shift of irradiated LWR-fuels at high burn-ups: Review of data and alternative interpretation of recently published results

International Nuclear Information System (INIS)

Spino, J.; Peerani, P.

2008-01-01

The available oxygen potential data of LWR-fuels by the EFM-method have been reviewed and compared with thermodynamic data of equivalent simulated fuels and mixed oxide systems, combined with the analysis of lattice parameter data. Up to burn-ups of 70-80 GWd/tM the comparison confirmed traditional predictions anticipating the fuels to remain quasi stoichiometric along irradiation. However, recent predictions of a fuel with average burn-up around 100 GWd/tM becoming definitely hypostoichiometric were not confirmed. At average burn-ups around 80 GWd/tM and above, it is shown that the fuels tend to acquire progressively slightly hyperstoichiometric O/M ratios. The maximum derived O/M ratio for an average burn-up of 100 GWd/tM lies around 2.001 and 2.002. Though slight, the stoichiometry shift may have a measurable accelerating impact on fission gas diffusion and release

Cellular automata approach to investigation of high burn-up structures in nuclear reactor fuel

International Nuclear Information System (INIS)

Akishina, E.P.; Ivanov, V.V.; Kostenko, B.F.

2005-01-01

Micrographs of uranium dioxide (UO 2 ) corresponding to exposure times in reactor during 323, 953, 971, 1266 and 1642 full power days were investigated. The micrographs were converted into digital files isomorphous to cellular automata (CA) checkerboards. Such a representation of the fuel structure provides efficient tools for its dynamics simulation in terms of primary 'entities' imprinted in the micrographs. Besides, it also ensures a possibility of very effective micrograph processing by CA means. Interconnection between the description of fuel burn-up development and some exactly soluble models is ascertained. Evidences for existence of self-organization in the fuel at high burn-ups were established. The fractal dimension of microstructures is found to be an important characteristic describing the degree of radiation destructions

Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn

International Nuclear Information System (INIS)

Talamo, Alberto; Ji, Wei; Cetnar, Jerzy; Gudowski, Waclaw

2006-01-01

Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides

Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Royal Institute of Technology (KTH), Roslagstullsbacken 21, Stockholm S-10691 (Sweden)]. E-mail: alby@anl.gov; Ji, Wei [University of Michigan, Bonisteel Boulevard 2355, Ann Arbor, MI 48109-2104 (United States); Cetnar, Jerzy [AGH-University of Science and Technology, Al. Mickiewicza 30 Cracow (Poland); Gudowski, Waclaw [Royal Institute of Technology (KTH), Roslagstullsbacken 21, Stockholm S-10691 (Sweden)

2006-09-15

Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides.

Fuel element burnup determination in HEU-LEU mixed TRIGA research reactor core

International Nuclear Information System (INIS)

Zagar, Tomaz; Ravnik, Matjaz

2000-01-01

This paper presents the results of a burnup calculations and burnup measurements for TRIGA FLIP HEU fuel elements and standard TRIGA LEU fuel elements used simultaneously in small TRIGA Mark II research reactor in Ljubljana, Slovenija. The fuel element burnup for approximately 15 years of operation was calculated with two different in house computer codes TRIGAP and TRIGLAV (both codes are available at OECD NEA Data Bank). The calculation is performed in one-dimensional radial geometry in TRIGAP and in two-dimensional (r,φ) geometry in TRIGLAV. Inter-comparison of results shows important influence of in-core water gaps, irradiation channels and mixed rings on burnup calculation accuracy. Burnup of 5 HEU and 27 LEU fuel elements was also measured with reactivity method. Measured and calculated burnup values are inter-compared for these elements (author)

Plasma nitriding of steels

CERN Document Server

Aghajani, Hossein

2017-01-01

This book focuses on the effect of plasma nitriding on the properties of steels. Parameters of different grades of steels are considered, such as structural and constructional steels, stainless steels and tools steels. The reader will find within the text an introduction to nitriding treatment, the basis of plasma and its roll in nitriding. The authors also address the advantages and disadvantages of plasma nitriding in comparison with other nitriding methods. .

Achieving High Burnup Targets With Mox Fuels: Techno Economic Implications

International Nuclear Information System (INIS)

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

2013-01-01

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

A simulation of the temperature overshoot observed at high burnup in annular fuel pellets

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

Instrumented experiments have been carried out in recent years to calibrate and improve temperature calculations at high burnup in PWR nuclear fuel rods. The introduction of a thermocouple in the fuel stack allows the experiment to record the centre-line temperature all along the irradiation or re-irradiation. The results obtained on fresh fuel have not revealed any abnormal behavior as have observations done on high burnup rods. In this case, a sudden overshoot has been recorded on the thermocouple temperature above an average power threshold. Several hypotheses have been suggested. Only two seem to be acceptable: one in relation to an effect of grain decohesion, another based on a modification of fuel chemistry. The apparent reversibility of the phenomena when power decreases led us to prefer the first explanation. Indeed, the introduction of a thermocouple means that annular fuel pellets must be used. These are either initially manufactured with a central hole or drilled after base irradiation, using the ``RISOE`` technique. One must bear in mind that the use of such annular pellets drastically changes the crack pattern as irradiation proceeds. This is due to a different stress field which, combined with a weakening of the grain binding energy, leads to a partial grain decohesion on the inner face of the annular pellet. Modification of the grain binding energy is related to the presence of an increasing local population of gas bubbles and metallic precipitates at grain boundaries, as swelling creates intergranular local stresses which also could probably enhance the grain decohesion process. This grain decohesion concerns a 250 to 350 {mu}m depth and shows a narrow cracks network through which released fission gas can flow, temporarily pushing the resident helium gas out. The low conductivity of these gaseous fission products and the numerous gas layers created this way could partly explain the unexpected temperatures measured in high burnup fuels. (Abstract

Burnup studies of the subcritical fusion-driven in-zinerator

International Nuclear Information System (INIS)

Persson, C. M.; Gudowski, W.; Venneri, F.

2007-01-01

A fusion-driven subcritical core, 'In-Zinerator', has been proposed for nuclear waste transmutation [1]. In this concept, a powerful Z-pinch neutron source will produce pulses of 14 MeV neutrons that multiply in a surrounding subcritical core consisting of spent fuel from the LWR fuel cycle or from deep burn high temperature reactors. The proposed design has pulse frequency 0.1 Hz and a thermal power of 3 GWth. The Z-pinch fusion experiment is located at Sandia Laboratories, USA, and can today fire once a day. However, investigations have been made how to increase the frequency to several fires per minute. Each fire yields 300 MJ corresponding to 1020 neutrons per pulse. The source chamber will in the In-Zinerator concept be surrounded by spent fuel to reach an effective multiplication factor, k e ff, of 0.97. The core will be cooled by liquid lead. In this paper, the burnup of different fuel compositions in the In-Zinerator will be studied as function of initial k e ff. The Monte Carlo based continuous energy burnup code MCB [2][3]will be used. References: [1] B.B. Cipiti, Fusion Transmutation of Waste and the Role of the In-Zinerator in the Nuclear Fuel Cycle, Sandia Report SAND2006-3522, Sandia National Laboratories, USA, 2006. [2] J. Cetnar, J Wallenius and W Gudowski, MCB: A continuous energy Monte-Carlo burnup simulation code, Actinide and fission product partitioning and transmutation, Proc. of the Fifth Int. Information Exchange Meeting, Mol, Belgium, 25-27 November 1998, 523, OECD/NEA, 1998. [3] http://www.nea.fr/abs/html/nea-1643.html

Burnup credit effect on proposed cask payloads

International Nuclear Information System (INIS)

Hall, I.K.

1989-01-01

The purpose of the Cask Systems Development Program (CSDP) is to develop a variety of cask systems which will allow safe and economical movement of commercial spent nuclear fuel and high-level waste from the generator to the Federal repository or Monitored Retrievable Storage (MRS) facility. Program schedule objectives for the initial phase of the CSDP include the development of certified spent fuel cask systems by 1995 to support Office of Civilian Radioactive Waste Management shipments from the utilities beginning in the late 1990s. Forty-nine proposals for developing a family of spent fuel casks were received and comparisons made. General conclusions that can be drawn from the comparisons are that (1) the new generation of casks will have substantially increased payloads in comparison to current casks, and (2) an even greater payload increase may be achievable with burnup credit. The ranges in the payload estimates do not allow a precise separation of the payload increase attributable to the proposed allowance of fuel burnup credit, as compared wilt the no-burnup-credit case. The beneficial effects of cask payload increases on overall costs and risks of transporting spent fuel are significant; therefore further work aimed toward taking advantage of burnup credit is warranted

Full Core Burn-up Calculation at JRR-3 with MVP-BURN

International Nuclear Information System (INIS)

Komeda, Masao; Yamamoto, Kazuyoshi; Kusunoki, Tsuyoshi

2008-01-01

Research reactors use a burnable poison to suppress an excess reactivity in the beginning of reactor lifetime. The JRR-3 (Japan Research Reactor No.3) has used cadmium wires of radius 0.02 cm as a burnable poison. This report describes burn-up calculations of plate fuel models and full core models with MVP-BURN, which is a burn-up calculation code using Monte Carlo method and has been developed in JAEA (Japan Atomic Energy Agency). As the results of calculations of plate models, between a model composed of one burn-up region along the radius direction and a model composed of a few burn-up regions along the radius direction, the effective absorption cross section of 113 Cd has had different tendency on reaching approximate 40. day (10000 MWd/t). And as results of calculations of full core model, it has been indicated that k eff is almost same till approximate 80. day (22000 MWd/t) between a model composed of one burn-up region along the vertical direction and a model composed of a few burn-up regions along the vertical direction. However difference of 113 Cd burn-up becomes pronounced and each k eff makes a difference after 80. day. (authors)

Nuclear fuels with high burnup: safety requirements

International Nuclear Information System (INIS)

Phuc Tran Dai

2016-01-01

Vietnam authorities foresees to build 3 reactors from Russian design (VVER AES 2006) by 2030. In order to prepare the preliminary report on safety analysis the Vietnamese Agency for Radioprotection and Safety has launched an investigation on the behaviour of nuclear fuels at high burnups (up to 60 GWj/tU) that will be those of the new plants. This study deals mainly with the behaviour of the fuel assemblies in case of loss of coolant (LOCA). It appears that for an average burnup of 50 GWj/tU and for the advanced design of the fuel assembly (cladding and materials) safety requirements are fulfilled. For an average burnup of 60 GWj/tU, a list of issues remains to be assessed, among which the impact of clad bursting or the hydrogen embrittlement of the advanced zirconium alloys. (A.C.)

Burnup credit in a dry storage module

International Nuclear Information System (INIS)

Thornton, J.R.

1989-01-01

Comparison of spent fuel storage expansion options available to Oconee Nuclear Station revealed that dry storage could be economically competitive with transshipment and rod consolidation. Economic competitiveness, however, mandated large unit capacity while existing cask handling facilities at Oconee severely limited size and weight. The dry storage concept determined to best satisfy these conflicting criteria is a 24 pressurized water reactor (PWR) fuel assembly capacity NUTECH Horizontal Modular Storage (NUHOMS) system. The Oconee version of the NUHOMS system takes advantage of burnup credit in demonstrating criticality safety. The burnup credit criticality analysis was performed by Duke Power Company's Design Engineering Department. This paper was prepared to summarize the criticality control design features employed in the Oconee NUHOMS-24P DSC basket and to describe the incentives for pursuing a burnup credit design. Principal criticality design parameters, criteria, and analysis methodology are also presented

Burnup measurements at the RECH-1 research reactor

International Nuclear Information System (INIS)

Henriquez, C.; Navarro, G.; Pereda, C.; Torres, H.; Pena, L.; Klein, J.; Calderon, D.; Kestelman, A.J.

2002-01-01

The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)

Application of burnup credit in spent fuel management at Russian NPPs

International Nuclear Information System (INIS)

Koulikov, V.I.; Makarchuk, T.F.; Tikhonov, N.S.

1998-01-01

The article concerns implementation of burnup credit in spent fuel storage and transportation. Some of the problems with increased enrichment fuel can be resolved by use of modified transport methodology. Such as shipping in gas-filled casks only, reduced number of assemblies in casks, etc. However, the use of modified schemes of transportation results in essential financial losses. An actinide-only burnup credit is taken into account in most part of criticality calculations, and a parameter limiting loading of spent fuel in the cask or the repository is the avenge value of burnup on an assembly. The main method of burnup depth definition is its defect measurement. A short description of devices for measurement as well as some technical results of suing burnup credit approach in storage and transport are given. (author)

Corrosion studies with high burnup light water reactor fuel. Release of nuclides into simulated groundwater during accumulated contact time of up to two years

Energy Technology Data Exchange (ETDEWEB)

Zwicky, Hans-Urs (Zwicky Consulting GmbH, Remigen (Switzerland)); Low, Jeanett; Ekeroth, Ella (Studsvik Nuclear AB, Nykoeping (Sweden))

2011-03-15

In the framework of comprehensive research work supporting the development of a Swedish concept for the disposal of highly radioactive waste and spent fuel, Studsvik has performed a significant number of spent fuel corrosion studies under a variety of different conditions. These experiments, performed between 1990 and 2002, covered a burnup range from 27 to 49 MWd/kgU, which was typical for fuel to be disposed at that time. As part of this work, the so called Series 11 tests were performed under oxidising conditions in synthetic groundwater with fuel samples from a rod irradiated in the Ringhals 1 Boiling Water Reactor (BWR). In the meantime, Swedish utilities tend to increase the discharge burnup of fuel operated in their reactors. This means that knowledge of spent fuel corrosion performance has to be extended to higher burnup as well. Therefore, a series of experiments has been started at Studsvik, aiming at extending the data base acquired in the Series 11 corrosion tests to higher burnup fuel. Fuel burnup leads to complex and significant changes in the composition and properties of the fuel. The transformed microstructure, which is referred to as the high burnup structure or rim structure in the outer region of the fuel, consists of small grains of submicron size and a high concentration of pores of typical diameter 1 to 2 mum. This structure forms in UO{sub 2} fuel at a local burnup above 50 MWd/kgU, as long as the temperature is below 1,000-1,100 deg C. The high burnup at the pellet periphery is the consequence of plutonium build-up by neutron capture in 238U followed by fission of the formed plutonium. The amount of fission products in the fuel increases more or less linearly with burnup, in contrast to alpha emitting actinides that increase above average. As burnup across a spent fuel pellet is not uniform, but increases towards the periphery, the radiation field is also larger at the pellet surface. At the same time, it is easier for water to access the

Corrosion studies with high burnup light water reactor fuel. Release of nuclides into simulated groundwater during accumulated contact time of up to two years

International Nuclear Information System (INIS)

Zwicky, Hans-Urs; Low, Jeanett; Ekeroth, Ella

2011-03-01

In the framework of comprehensive research work supporting the development of a Swedish concept for the disposal of highly radioactive waste and spent fuel, Studsvik has performed a significant number of spent fuel corrosion studies under a variety of different conditions. These experiments, performed between 1990 and 2002, covered a burnup range from 27 to 49 MWd/kgU, which was typical for fuel to be disposed at that time. As part of this work, the so called Series 11 tests were performed under oxidising conditions in synthetic groundwater with fuel samples from a rod irradiated in the Ringhals 1 Boiling Water Reactor (BWR). In the meantime, Swedish utilities tend to increase the discharge burnup of fuel operated in their reactors. This means that knowledge of spent fuel corrosion performance has to be extended to higher burnup as well. Therefore, a series of experiments has been started at Studsvik, aiming at extending the data base acquired in the Series 11 corrosion tests to higher burnup fuel. Fuel burnup leads to complex and significant changes in the composition and properties of the fuel. The transformed microstructure, which is referred to as the high burnup structure or rim structure in the outer region of the fuel, consists of small grains of submicron size and a high concentration of pores of typical diameter 1 to 2 μm. This structure forms in UO 2 fuel at a local burnup above 50 MWd/kgU, as long as the temperature is below 1,000-1,100 deg C. The high burnup at the pellet periphery is the consequence of plutonium build-up by neutron capture in 238 U followed by fission of the formed plutonium. The amount of fission products in the fuel increases more or less linearly with burnup, in contrast to alpha emitting actinides that increase above average. As burnup across a spent fuel pellet is not uniform, but increases towards the periphery, the radiation field is also larger at the pellet surface. At the same time, it is easier for water to access the

Burnup effect on nuclear fuel cycle cost using an equilibrium model

International Nuclear Information System (INIS)

Youn, S. R.; Kim, S. K.; Ko, W. I.

2014-01-01

The degree of fuel burnup is an important technical parameter to the nuclear fuel cycle, being sensitive and progressive to reduce the total volume of process flow materials and eventually cut the nuclear fuel cycle costs. This paper performed the sensitivity analysis of the total nuclear fuel cycle costs to changes in the technical parameter by varying the degree of burnups in each of the three nuclear fuel cycles using an equilibrium model. Important as burnup does, burnup effect was used among the cost drivers of fuel cycle, as the technical parameter. The fuel cycle options analyzed in this paper are three different fuel cycle options as follows: PWR-Once Through Cycle(PWR-OT), PWR-MOX Recycle, Pyro-SFR Recycle. These fuel cycles are most likely to be adopted in the foreseeable future. As a result of the sensitivity analysis on burnup effect of each three different nuclear fuel cycle costs, PWR-MOX turned out to be the most influenced by burnup changes. Next to PWR-MOX cycle, in the order of Pyro-SFR and PWR-OT cycle turned out to be influenced by the degree of burnup. In conclusion, the degree of burnup in the three nuclear fuel cycles can act as the controlling driver of nuclear fuel cycle costs due to a reduction in the volume of spent fuel leading better availability and capacity factors. However, the equilibrium model used in this paper has a limit that time-dependent material flow and cost calculation is impossible. Hence, comparative analysis of the results calculated by dynamic model hereafter and the calculation results using an equilibrium model should be proceed. Moving forward to the foreseeable future with increasing burnups, further studies regarding alternative material of high corrosion resistance fuel cladding for the overall

Analytical and Experimental Evaluation of Joining Silicon Carbide to Silicon Carbide and Silicon Nitride to Silicon Nitride for Advanced Heat Engine Applications Phase II

Energy Technology Data Exchange (ETDEWEB)

Sundberg, G.J.

1994-01-01

Techniques were developed to produce reliable silicon nitride to silicon nitride (NCX-5101) curved joins which were used to manufacture spin test specimens as a proof of concept to simulate parts such as a simple rotor. Specimens were machined from the curved joins to measure the following properties of the join interlayer: tensile strength, shear strength, 22 C flexure strength and 1370 C flexure strength. In parallel, extensive silicon nitride tensile creep evaluation of planar butt joins provided a sufficient data base to develop models with accurate predictive capability for different geometries. Analytical models applied satisfactorily to the silicon nitride joins were Norton's Law for creep strain, a modified Norton's Law internal variable model and the Monkman-Grant relationship for failure modeling. The Theta Projection method was less successful. Attempts were also made to develop planar butt joins of siliconized silicon carbide (NT230).

Revised SWAT. The integrated burnup calculation code system

International Nuclear Information System (INIS)

Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide

2000-07-01

SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

Revised SWAT. The integrated burnup calculation code system

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)

2000-07-01

SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

Conceptual cask design with burnup credit

International Nuclear Information System (INIS)

Lee, Seong Hee; Ahn, Joon Gi; Hwang, Hae Ryong

2003-01-01

Conceptual design has been performed for a spent fuel transport cask with burnup credit and a neutron-absorbing material to maximize transportation capacity. Both fresh and burned fuel are assumed to be stored in the cask and boral and borated stainless steel are selected for the neutron-absorbing materials. Three different sizes of cask with typical 14, 21 and 52 PWR fuel assemblies are modeled and analyzed with the SCALE 4.4 code system. In this analysis, the biases and uncertainties through validation calculations for both isotopic predictions and criticality calculation for the spent fuel have been taken into account. All of the reactor operating parameters, such as moderator density, soluble boron concentration, fuel temperature, specific power, and operating history, have been selected in a conservative way for the criticality analysis. Two different burnup credit loading curves are developed for boral and borated stainless steel absorbing materials. It is concluded that the spent fuel transport cask design with burnup credit is feasible and is expected to increase cask payloads. (author)

Use of burnup credit for transportation and storage

International Nuclear Information System (INIS)

Sanders, T.L.; Ewing, R.I.; Lake, W.H.

1991-01-01

Burnup credit is the application of the effects of fuel burnup to nuclear criticality design. When burnup credit is considered in the design of storage facilities and transportation casks for spent fuel, the objectives are to reduce the requirements for storage space and to increase the payload of casks with acceptable nuclear criticality safety margins. The spent-fuel carrying capacities of previous-generation transport casks have been limited primarily by requirements to remove heat and/or to provide shielding. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced significantly. Thus a considerable weight margin is available to the designer for increasing the payload capacity. One method to achieve an increase in capacity is to reduce fuel assembly spacing. The amount of reduction in assembly spacing is limited by criticality and fuel support structural concerns. The optimum fuel assembly spacing provides the maximum cask loading within a basket that has adequate criticality control and sufficient structural integrity for regulatory accident scenarios. The incorporation of burnup credit in cask designs could result in considerable benefits in the transport of spent fuel. The acceptance of burnup credit for the design of transport casks depends on the resolution of system safety issues and the uncertainties that affect the determination of criticality safety margins. The remainder of this report will examine these issues and the integrated approach under way to resolve them. 20 refs., 2 figs

CHAR and BURNMAC - burnup modules of the AUS neutronics code system

International Nuclear Information System (INIS)

Robinson, G.S.

1986-03-01

In the AUS neutronics code system, the burnup module CHAR solves the nuclide depletion equations by an analytic technique in a number of spatial zones. CHAR is usually used as one component of a lattice burnup calculation but contains features which also make it suitable for some global burnup calculations. BURNMAC is a simple accounting module based on the assumption that cross sections for a rector zone depend only on irradiation. BURNMAC is used as one component of a global calculation in which burnup is achieved by interpolation in the cross sections produced from a previous lattice calculation

Cell verification of parallel burnup calculation program MCBMPI based on MPI

International Nuclear Information System (INIS)

Yang Wankui; Liu Yaoguang; Ma Jimin; Wang Guanbo; Yang Xin; She Ding

2014-01-01

The parallel burnup calculation program MCBMPI was developed. The program was modularized. The parallel MCNP5 program MCNP5MPI was employed as neutron transport calculation module. And a composite of three solution methods was used to solve burnup equation, i.e. matrix exponential technique, TTA analytical solution, and Gauss Seidel iteration. MPI parallel zone decomposition strategy was concluded in the program. The program system only consists of MCNP5MPI and burnup subroutine. The latter achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, and data exchanging with MCNP5MPI. Also, the program was verified with the pressurized water reactor (PWR) cell burnup benchmark. The results show that it,s capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)

End effect Keff bias curve for actinide-only burnup credit casks

International Nuclear Information System (INIS)

Kang, C.H.; Lancaster, D.B.

1997-01-01

A conservative end effect k eff bias curve for actinide-only burnup credit for spent fuel casks is presented in this paper. The k eff bias values can be added to the uniform axial burnup analysis to conservatively bound the actinide-only end effect. A normalized axial burnup distribution for the standard Westinghouse 17 x 17 assembly design is used for calculating k eff . The end effect calculated is a strong function of burnup, and increases as cask size size decreases. The presence of poison plates increases the end effect. The bias curve presented is based on the most limiting cask configuration of a single PWR assembly with completely black poison plates. Therefore, axially uniform criticality calculations with application of the proposed k eff could eliminate the need for axially burnup dependent analyses. 7 refs., 1 fig

Theory analysis and simple calculation of travelling wave burnup scheme

International Nuclear Information System (INIS)

Zhang Jian; Yu Hong; Gang Zhi

2012-01-01

Travelling wave burnup scheme is a new burnup scheme that breeds fuel locally just before it burns. Based on the preliminary theory analysis, the physical imagine was found. Through the calculation of a R-z cylinder travelling wave reactor core with ERANOS code system, the basic physical characteristics of this new burnup scheme were concluded. The results show that travelling wave reactor is feasible in physics, and there are some good features in the reactor physics. (authors)

Effect of error propagation of nuclide number densities on Monte Carlo burn-up calculations

International Nuclear Information System (INIS)

Tohjoh, Masayuki; Endo, Tomohiro; Watanabe, Masato; Yamamoto, Akio

2006-01-01

As a result of improvements in computer technology, the continuous energy Monte Carlo burn-up calculation has received attention as a good candidate for an assembly calculation method. However, the results of Monte Carlo calculations contain the statistical errors. The results of Monte Carlo burn-up calculations, in particular, include propagated statistical errors through the variance of the nuclide number densities. Therefore, if statistical error alone is evaluated, the errors in Monte Carlo burn-up calculations may be underestimated. To make clear this effect of error propagation on Monte Carlo burn-up calculations, we here proposed an equation that can predict the variance of nuclide number densities after burn-up calculations, and we verified this equation using enormous numbers of the Monte Carlo burn-up calculations by changing only the initial random numbers. We also verified the effect of the number of burn-up calculation points on Monte Carlo burn-up calculations. From these verifications, we estimated the errors in Monte Carlo burn-up calculations including both statistical and propagated errors. Finally, we made clear the effects of error propagation on Monte Carlo burn-up calculations by comparing statistical errors alone versus both statistical and propagated errors. The results revealed that the effects of error propagation on the Monte Carlo burn-up calculations of 8 x 8 BWR fuel assembly are low up to 60 GWd/t

Burnup credit demands for spent fuel management in Ukraine

International Nuclear Information System (INIS)

Medun, V.

2001-01-01

In fact, till now, burnup credit has not be applied in Ukrainian nuclear power for spent fuel management systems (storage and transport). However, application of advanced fuel at VVER reactors, arising spent fuel amounts, represent burnup credit as an important resource to decrease spent fuel management costs. The paper describes spent fuel management status in Ukraine from viewpoint of subcriticality assurance under spent fuel storage and transport. It also considers: 1. Regulation basis concerning subcriticality assurance, 2. Basic spent fuel and transport casks characteristics, 3. Possibilities and demands for burnup credit application at spent fuel management systems in Ukraine. (author)

Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

Energy Technology Data Exchange (ETDEWEB)

Enercon Services, Inc.

2011-03-14

Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in

Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

International Nuclear Information System (INIS)

2011-01-01

Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in

Inter-layer potential for hexagonal boron nitride

Science.gov (United States)

Leven, Itai; Azuri, Ido; Kronik, Leeor; Hod, Oded

2014-03-01

A new interlayer force-field for layered hexagonal boron nitride (h-BN) based structures is presented. The force-field contains three terms representing the interlayer attraction due to dispersive interactions, repulsion due to anisotropic overlaps of electron clouds, and monopolar electrostatic interactions. With appropriate parameterization, the potential is able to simultaneously capture well the binding and lateral sliding energies of planar h-BN based dimer systems as well as the interlayer telescoping and rotation of double walled boron-nitride nanotubes of different crystallographic orientations. The new potential thus allows for the accurate and efficient modeling and simulation of large-scale h-BN based layered structures.

Inter-layer potential for hexagonal boron nitride

Energy Technology Data Exchange (ETDEWEB)

Leven, Itai; Hod, Oded, E-mail: odedhod@tau.ac.il [Department of Chemical Physics, School of Chemistry, The Raymond and Beverly Sackler Faculty of Exact Sciences, Tel-Aviv University, Tel-Aviv 69978 (Israel); Azuri, Ido; Kronik, Leeor [Department of Materials and Interfaces, Weizmann Institute of Science, Rehovoth 76100 (Israel)

2014-03-14

A new interlayer force-field for layered hexagonal boron nitride (h-BN) based structures is presented. The force-field contains three terms representing the interlayer attraction due to dispersive interactions, repulsion due to anisotropic overlaps of electron clouds, and monopolar electrostatic interactions. With appropriate parameterization, the potential is able to simultaneously capture well the binding and lateral sliding energies of planar h-BN based dimer systems as well as the interlayer telescoping and rotation of double walled boron-nitride nanotubes of different crystallographic orientations. The new potential thus allows for the accurate and efficient modeling and simulation of large-scale h-BN based layered structures.

Inter-layer potential for hexagonal boron nitride

International Nuclear Information System (INIS)

Leven, Itai; Hod, Oded; Azuri, Ido; Kronik, Leeor

2014-01-01

A new interlayer force-field for layered hexagonal boron nitride (h-BN) based structures is presented. The force-field contains three terms representing the interlayer attraction due to dispersive interactions, repulsion due to anisotropic overlaps of electron clouds, and monopolar electrostatic interactions. With appropriate parameterization, the potential is able to simultaneously capture well the binding and lateral sliding energies of planar h-BN based dimer systems as well as the interlayer telescoping and rotation of double walled boron-nitride nanotubes of different crystallographic orientations. The new potential thus allows for the accurate and efficient modeling and simulation of large-scale h-BN based layered structures

Fuel cycle cost considerations of increased discharge burnups

International Nuclear Information System (INIS)

Scherpereel, L.R.; Frank, F.J.

1982-01-01

Evaluations are presented that indicate the attainment of increased discharge burnups in light water reactors will depend on economic factors particular to individual operators. In addition to pure resource conserving effects and assuming continued reliable fuel performance, a substantial economic incentive must exist to justify the longer operating times necessary to achieve higher burnups. Whether such incentive will exist or not will depend on relative price levels of all fuel cycle cost components, utility operating practices, and resolution of uncertainties associated with the back-end of the fuel cycle. It is concluded that implementation of increased burnups will continue at a graduated pace similar to past experience, rather than finding universal acceptance of particular increased levels at any particular time

Fission gas release from fuels at high burnup

International Nuclear Information System (INIS)

Kauffmann, Yves; Pointud, M.L.; Vignesoult, Nicole; Atabek, Rosemarie; Baron, Daniel.

1982-04-01

Determinations of residual gas concentrations by heating and by X microanalysis were respectively carried out on particles (TANGO program) and on sections of fuel rods, perfectly characterized as to fabrication and irradiation history. A threshold release temperature of 1250 0 C+-100 0 C was determined irrespective of the type of oxide and the irradiation history in the 18,000-45,000 MWdt -1 (U) specific burnup field. The overall analyses of gas released from the fuel rods show that, in the PWR operating conditions, the fraction released remains less than 1% up to a mean specific burnup of 35000 MWdt -1 (U). The release of gases should not be a limiting factor in the increase of specific burnups [fr

Monte Carlo burnup codes acceleration using the correlated sampling method

International Nuclear Information System (INIS)

Dieudonne, C.

2013-01-01

For several years, Monte Carlo burnup/depletion codes have appeared, which couple Monte Carlo codes to simulate the neutron transport to deterministic methods, which handle the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3-dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the Monte Carlo solver called at each time step. In this document we present an original methodology to avoid the repetitive and time-expensive Monte Carlo simulations, and to replace them by perturbation calculations: indeed the different burnup steps may be seen as perturbations of the isotopic concentration of an initial Monte Carlo simulation. In a first time we will present this method, and provide details on the perturbative technique used, namely the correlated sampling. In a second time we develop a theoretical model to study the features of the correlated sampling method to understand its effects on depletion calculations. In a third time the implementation of this method in the TRIPOLI-4 code will be discussed, as well as the precise calculation scheme used to bring important speed-up of the depletion calculation. We will begin to validate and optimize the perturbed depletion scheme with the calculation of a REP-like fuel cell depletion. Then this technique will be used to calculate the depletion of a REP-like assembly, studied at beginning of its cycle. After having validated the method with a reference calculation we will show that it can speed-up by nearly an order of magnitude standard Monte-Carlo depletion codes. (author) [fr

Assessment of US NRC fuel rod behavior codes to extended burnup

International Nuclear Information System (INIS)

Laats, E.T.; Croucher, D.W.; Haggag, F.M.

1982-01-01

The purpose of this paper is to report the status of assessing the capabilities of the NRC fuel rod performance codes for calculating extended burnup rod behavior. As part of this effort, a large spectrum of fuel rod behavior phenomena was examined, and the phenomena deemed as being influential during extended burnup operation were identified. Then, the experiment data base addressing these identified phenomena was examined for availability and completeness at extended burnups. Calculational capabilities of the NRC's steady state FRAPCON-2 and transient FRAP-T6 fuel rod behavior codes were examined for each of the identified phenomenon. Parameters calculated by the codes were compared with the available data base, and judgments were made regarding model performance. Overall, the FRAPCON-2 code was found to be moderately well assessed to extended burnups, but the FRAP-T6 code cannot be adequately assessed until more transient high burnup data are available

Triton burnup measurements in KSTAR using a neutron activation system

Energy Technology Data Exchange (ETDEWEB)

Jo, Jungmin; Shi, Yue-Jiang; Chung, Kyoung-Jae, E-mail: jkjlsh1@snu.ac.k; Hwang, Y. S. [Department of Nuclear Engineering, Seoul National University, Seoul 151-744 (Korea, Republic of); Cheon, MunSeong; Rhee, T.; Kim, Junghee [National Fusion Research Institute, Daejeon 34133 (Korea, Republic of); Kim, Jun Young [Korea University of Science and Technology, Daejeon 34133 (Korea, Republic of); Isobe, M.; Ogawa, K. [National Institute for Fusion Science, Toki-shi (Japan); SOKENDAI (The Graduate University for Advanced Studies), Toki-shi (Japan)

2016-11-15

Measurements of the time-integrated triton burnup for deuterium plasma in Korea Superconducting Tokamak Advanced Research (KSTAR) have been performed following the simultaneous detection of the d-d and d-t neutrons. The d-d neutrons were measured using a {sup 3}He proportional counter, fission chamber, and activated indium sample, whereas the d-t neutrons were detected using activated silicon and copper samples. The triton burnup ratio from KSTAR discharges is found to be in the range 0.01%–0.50% depending on the plasma conditions. The measured burnup ratio is compared with the prompt loss fraction of tritons calculated with the Lorentz orbit code and the classical slowing-down time. The burnup ratio is found to increase as plasma current and classical slowing-down time increase.

Comparison of measured and calculated burn-up of AVR-Fuel-Elements

Energy Technology Data Exchange (ETDEWEB)

Wagemann, R.

1974-03-15

Burn-up comparisons are made for small batches of three types of AVR fuel elements using a coupled EREBUS-MUPO neutronic analysis compared against test results from both nondestructive gamma-ray measurements of cesium-137 activity and destructive mass spectrometry measurements of the ratio of U-233 to U-235. The comparisons are relatively good for average burn-up and reasonably good for burn-up distributions.

Ion nitriding of aluminium

International Nuclear Information System (INIS)

Fitz, T.

2002-09-01

The present study is devoted to the investigation of the mechanism of aluminium nitriding by a technique that employs implantation of low-energy nitrogen ions and diffusional transport of atoms. The nitriding of aluminium is investigated, because this is a method for surface modification of aluminium and has a potential for application in a broad spectrum of fields such as automobile, marine, aviation, space technologies, etc. However, at present nitriding of aluminium does not find any large scale industrial application, due to problems in the formation of stoichiometric aluminium nitride layers with a sufficient thickness and good quality. For the purposes of this study, ion nitriding is chosen, as an ion beam method with the advantage of good and independent control over the process parameters, which thus can be related uniquely to the physical properties of the resulting layers. Moreover, ion nitriding has a close similarity to plasma nitriding and plasma immersion ion implantation, which are methods with a potential for industrial application. (orig.)

Triton burnup in JET

International Nuclear Information System (INIS)

Chipsham, E.; Jarvis, O.N.; Sadler, G.

1989-01-01

Triton burnup measurements have been made at JET using time-integrated copper activation and time-resolved silicon detector techniques. The results confirm the classical nature of both the confinement and the slowing down of the 1 MeV tritons in a plasma. (author) 8 refs., 3 figs

Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

International Nuclear Information System (INIS)

Hussein, H.M.; Sakr, A.M.; Amin, E.H.

2011-01-01

Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

Burnup credit implementation plan and preparation work at JAERI

International Nuclear Information System (INIS)

Nomura, Y.; Itahara, K.

2001-01-01

Application of the burnup credit concept is considered to be very effective to the design of spent fuel transport and storage facilities. This technology is all the more important when considering construction of the intermediate spent fuel storage facility, which is to be commissioned by 2010 due to increasing amount of accumulated spent fuel in Japan. Until reprocessing and recycling all the spent fuel arising, they will be stored as an energy stockpile until such time as they can be reprocessed. On the other hand, the burnup credit has been partly taken into account for the spent fuel management at Rokkasho Reprocessing Plant, which is to be commissioned in 2005. They have just finished the calibration tests for their burnup monitor with initially accepted several spent fuel assemblies. Because this monitoring system is employed with highly conservative safety margin, it is considered necessary to develop the more rational and simplified method to confirm burnup of spent fuel. A research program has been instituted to improve the present method employed at the spent fuel management system for the Spent Fuel Receiving and Storage Pool of Rokkasho Reprocessing Plant. This program is jointly performed by Japan Nuclear Fuel Limited (JNFL) and JAERI.This presentation describes the current status of spent fuel accumulation discharged from PWR and BWR in Japan and the recent incentive to introduce burnup credit into design of spent fuel storage and transport facilities. This also includes the content of the joint research program initiated by JNFL and JAERI. The relevant study has been continued at JAERI. The results by these research programs will be included in the Burnup Credit Guide Original Version compiled by JAERI. (author)

Determination of enrichment of recycle uranium fuels for different burnup values

International Nuclear Information System (INIS)

Zabunoglu, Okan H.

2008-01-01

Uranium (U) recovered from spent LWR fuels by reprocessing, which contains small amounts of U-236, is to be enriched before being re-irradiated as the recycle U. During the enrichment of recovered U in U-235, the mass fraction of U-236 also increases. Since the existence of U-236 in the recycle U has a negative effect on neutron economy, a greater enrichment of U-235 in the recycle U is required for reaching the same burnup as can be reached by the fresh U fuel. Two burnup values play the most important role in determining the enrichment of recycle U: (1) discharge burnup of spent fuel from which the recycle U is obtained and (2) desired discharge burnup of the recycle U fuel. A step-by-step procedure for calculating the enrichment of the recycle U as a function of these two burnup values is introduced. The computer codes MONTEBURNS and ORIGEN-S are made use of and a three-component (U-235, U-236, U-238) enrichment scheme is applied for calculating the amount of U-236 in producing the recycle U from the recovered U. As was aimed, the resulting expression is simple enough for quick/hand calculations of the enrichment of the recycle U for any given discharge burnup of spent fuel and for any desired discharge burnup of the recycle U fuel, most accurately within the range of 33,000-50,000 MWd/tonU

Increased fuel burn-up and fuel cycle equilibrium

International Nuclear Information System (INIS)

Debes, M.

2001-01-01

Improvement of nuclear competitiveness will rely mainly on increased fuel performance, with higher burn-up, and reactors sustained life. Regarding spent fuel management, the EDF current policy relies on UO 2 fuel reprocessing (around 850 MTHM/year at La Hague) and MOX recycling to ensure plutonium flux adequacy (around 100 MTHM/year, with an electricity production equivalent to 30 TWh). This policy enables to reuse fuel material, while maintaining global kWh economy with existing facilities. It goes along with current perspective to increase fuel burn-up up to 57 GWday/t mean in 2010. The following presentation describes the consequences of higher fuel burn-up on fuel cycle and waste management and implementation of a long term and global equilibrium for decades in spent fuel management resulting from this strategy. (author)

Triton burnup in JET - profile effects

Energy Technology Data Exchange (ETDEWEB)

Jarvis, O.N.; Conroy, S.W.; Marcus, F.B.; Sadler, G.J.; Belle, P. van (Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking); Adams, J.M.; Watkins, N. (AEA Industrial Technology, Harwell Laboratory (United Kingdom))

1991-01-01

Measurements of the 14 MeV neutron emission from triton burnup show that the 14 MeV emission profile shadows closely the 2,5 MeV profile but after a delay corresponding to the triton slowing down time. The slightly greater width of the 14 MeV neutron profile is a consequence of the finite Larmor radius of the tritons. It has not so far been possible to identify unambiguously any effects on the triton burnup that are attributable to sawtooth crashes. Finally, the time dependence of the triton profile indicates that the triton diffusion coefficient is very small (<<0.1 m[sup 2]/s). (author) 4 refs., 3 figs.

Triton burnup in JET - profile effects

International Nuclear Information System (INIS)

Jarvis, O.N.; Conroy, S.W.; Marcus, F.B.; Sadler, G.J.; Belle, P. van

1991-01-01

Measurements of the 14 MeV neutron emission from triton burnup show that the 14 MeV emission profile shadows closely the 2,5 MeV profile but after a delay corresponding to the triton slowing down time. The slightly greater width of the 14 MeV neutron profile is a consequence of the finite Larmor radius of the tritons. It has not so far been possible to identify unambiguously any effects on the triton burnup that are attributable to sawtooth crashes. Finally, the time dependence of the triton profile indicates that the triton diffusion coefficient is very small ( 2 /s). (author) 4 refs., 3 figs

Effect of local burn-up variation on computed mean nuclide concentrations

International Nuclear Information System (INIS)

Moeller, W.

1982-01-01

Mean concentrations of U-235, U-236, U-238, Pu-239, Pu-240, Pu-241 and Pu-242 in some volume areas of WWER-440 fuel assemblies have been calculated from corresponding burn-up microdistribution data and compared with those calculated from burn-up mean values. Differences occurring were below 3% for the uranium nuclides but, at low burn-ups, considerable for Pu-241 and Pu-242. (author)

Review of high burn-up RIA and LOCA database and criteria

International Nuclear Information System (INIS)

Vitanza, C.; Hrehor, M.

2006-01-01

This document is intended to provide regulators, their technical support organizations and industry with a concise review of existing fuel experimental data at RIA and LOCA conditions and considerations on how these data affect fuel safety criteria at increasing burn-up. It mostly addresses experimental results relevant to BWR and PWR fuel and it encompasses several contributions from the various experts that participated in the CSNI SEGFSM activities. It also covers the information presented at the joint CSNI/CNRA Topical Discussion on high burn-up fuel issues that took place on this subject in December 2004. The report is organized in the following way: the CABRI RIA database (14 tests), the NSRR database (26 tests) and other databases, RIA failure thresholds, comparison of failure thresholds for the HZP case, LOCA database ductility tests and quench tests, LOCA safety limit, provisional burn-up dependent criterion for Zr-4. The conclusions are as follows. On RIA, there is a well-established testing method and a significant and relatively consistent database from NSRR and Cabri tests, especially on high burn-up Zr-2 and Zr-4 cladding. It is encouraging that several correlations have been proposed for the RIA fuel failure threshold. Their predictions are compared and discussed in this paper for a representative PWR case. On LOCA, there are two different test methods, one based on ductility determinations and the other based on 'integral' quench tests. The LOCA database at high burn-up is limited to both testing methods. Ductility tests carried out with pre-hydrided non-irradiated cladding show a pronounced hydrogen effect. Data for actual high burn-up specimens are being gathered in various laboratories and will form the basis for a burn-up dependent LOCA limit. A provisional burn-up dependent criterion is discussed in the paper

Isotopic biases for actinide-only burnup credit

International Nuclear Information System (INIS)

Rahimi, M.; Lancaster, D.; Hoeffer, B.; Nichols, M.

1997-01-01

The primary purpose of this paper is to present the new methodology for establishing bias and uncertainty associated with isotopic prediction in spent fuel assemblies for burnup credit analysis. The analysis applies to the design of criticality control systems for spent fuel casks. A total of 54 spent fuel samples were modeled and analyzed using the Shielding Analyses Sequence (SAS2H). Multiple regression analysis and a trending test were performed to develop isotopic correction factors for 10 actinide burnup credit isotopes. 5 refs., 1 tab

Experimental studies of spent fuel burn-up in WWR-SM reactor

Energy Technology Data Exchange (ETDEWEB)

Alikulov, Sh. A.; Baytelesov, S.A.; Boltaboev, A.F.; Kungurov, F.R. [Institute of Nuclear Physics, Ulughbek township, 100214, Tashkent (Uzbekistan); Menlove, H.O.; O’Connor, W. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Osmanov, B.S., E-mail: bari_osmanov@yahoo.com [Research Institute of Applied Physics, Vuzgorodok, 100174 Tashkent (Uzbekistan); Salikhbaev, U.S. [Institute of Nuclear Physics, Ulughbek township, 100214, Tashkent (Uzbekistan)

2014-10-01

Highlights: • Uranium burn-up measurement from {sup 137}Cs activity in spent reactor fuel. • Comparison to reference sample with known burn-up value (ratio method). • Cross-check of the approach with neutron-based measurement technique. - Abstract: The article reports the results of {sup 235}U burn-up measurements using {sup 137}Cs activity technique for 12 nuclear fuel assemblies of WWR-SM research reactor after 3-year cooling time. The discrepancy between the measured and the calculated burn-up values was about 3%. To increase the reliability of the data and for cross-check purposes, neutron measurement approach was also used. Average discrepancy between two methods was around 12%.

A technique of melting temperature measurement and its application for irradiated high-burnup MOX fuels

International Nuclear Information System (INIS)

Namekawa, Takashi; Hirosawa, Takashi

1999-01-01

A melting temperature measurement technique for irradiated oxide fuels is described. In this technique, the melting temperature was determined from a thermal arrest on a heating curve of the specimen which was enclosed in a tungsten capsule to maintain constant chemical composition of the specimen during measurement. The measurement apparatus was installed in an alpha-tight steel box within a gamma-shielding cell and operated by remote handling. The temperature of the specimen was measured with a two-color pyrometer sighted on a black-body well at the bottom of the tungsten capsule. The diameter of the black-body well was optimized so that the uncertainties of measurement were reduced. To calibrate the measured temperature, two reference melting temperature materials, tantalum and molybdenum, were encapsulated and run before and after every oxide fuel test. The melting temperature data on fast reactor mixed oxide fuels irradiated up to 124 GWd/t were obtained. In addition, simulated high-burnup mixed oxide fuel up to 250 GWd/t by adding non-radioactive soluble fission products was examined. These data shows that the melting temperature decrease with increasing burnup and saturated at high burnup region. (author)

Safety aspects related to burnup increase and mixed oxide fuel

International Nuclear Information System (INIS)

Thomas, W.

1992-01-01

The dominant factor presently limiting the fuel burnup is the response of the cladding hulls. To maintain the excellent record of very low fuel failure rates for increased burnups further technical development is underway and necessary. In the nuclear fuel cycle increased burnups lead to a remarkable reduction of spent fuel arisings and corresponding economic savings. Thermal recycling of plutonium presently provides an opportunity to reduce the rising accumulation of plutunium in a situation where there is no demand for this fissile material in Fast Breeder Reactors. (orig.) [de

PIE and separate effect test of high burnup UO2 fuel

International Nuclear Information System (INIS)

Yang, Yong Sik; Kim, S.K.; Kim, D.H.

2005-01-01

To investigate the performance of a high burnup UO 2 fuel, the highest burnup fuel assembly in KOREA was transported to the PIE facility in KAERI. It was a 17·17 fuel assembly irradiated at the Ulchin Unit 2 PWR. The peak fuel rod average burnup was about 57MWd/kgU and locally 65MWd/kgU. The general PIE was performed to investigate the fuel rod irradiation performance. Fission gas release, burnup, oxide thickness, hydrogen pickup, CRUD, and density change were measured by destructive of non-destructive test. Microstructure change, bubble and pore size distributions were observed by optical microscopy, SEM and EPMA. All generated and available PIE results were used to verify high burnup fuel performance code INFRA. Several rods were cut for additional separate effect test. For the high burnup fission gas release behaviour analysis, annealing apparatus were developed and installed in hot cell and preliminary test was performed. In addition to current apparatus new induction furnace will be installed in hot cell to investigate the high temperature and transient fission gas release behaviour. Ring tensile test was performed to analyze the material property degradation which caused by the oxidation and hydride, and additional mechanical tests will be performed. (Author)

Reactivity effect of spent fuel due to spatial distributions for coolant temperature and burnup

Energy Technology Data Exchange (ETDEWEB)

Hayashi, T.; Yamane, Y. [Nagoya Univ., Dept. of Nuclear Engineering, Nagoya, Aichi (Japan); Suyama, K. [OECD/NEA, Paris (France); Mochizuki, H. [Japan Research Institute, Ltd., Tokyo (Japan)

2002-03-01

We investigated the reactivity effect of spent fuel caused by the spatial distributions of coolant temperature and burnup by using the integrated burnup calculation code system SWAT. The reactivity effect which arises from taking account of the spatial coolant temperature distribution increases as the average burnup increases, and reaches the maximum value of 0.69%{delta}k/k at 50 GWd/tU when the burnup distribution is concurrently considered. When the burnup distribution is ignored, the reactivity effect decreases by approximately one-third. (author)

Optimum burnup of BAEC TRIGA research reactor

International Nuclear Information System (INIS)

Lyric, Zoairia Idris; Mahmood, Mohammad Sayem; Motalab, Mohammad Abdul; Khan, Jahirul Haque

2013-01-01

Highlights: ► Optimum loading scheme for BAEC TRIGA core is out-to-in loading with 10 fuels/cycle starting with 5 for the first reload. ► The discharge burnup ranges from 17% to 24% of U235 per fuel element for full power (3 MW) operation. ► Optimum extension of operating core life is 100 MWD per reload cycle. - Abstract: The TRIGA Mark II research reactor of BAEC (Bangladesh Atomic Energy Commission) has been operating since 1986 without any reshuffling or reloading yet. Optimum fuel burnup strategy has been investigated for the present BAEC TRIGA core, where three out-to-in loading schemes have been inspected in terms of core life extension, burnup economy and safety. In considering different schemes of fuel loading, optimization has been searched by only varying the number of fuels discharged and loaded. A cost function has been defined and evaluated based on the calculated core life and fuel load and discharge. The optimum loading scheme has been identified for the TRIGA core, the outside-to-inside fuel loading with ten fuels for each cycle starting with five fuels for the first reload. The discharge burnup has been found ranging from 17% to 24% of U235 per fuel element and optimum extension of core operating life is 100 MWD for each loading cycle. This study will contribute to the in-core fuel management of TRIGA reactor

Performance analysis and simulation of vertical gallium nitride nanowire transistors

Science.gov (United States)

Witzigmann, Bernd; Yu, Feng; Frank, Kristian; Strempel, Klaas; Fatahilah, Muhammad Fahlesa; Schumacher, Hans Werner; Wasisto, Hutomo Suryo; Römer, Friedhard; Waag, Andreas

2018-06-01

Gallium nitride (GaN) nanowire transistors are analyzed using hydrodynamic simulation. Both p-body and n-body devices are compared in terms of threshold voltage, saturation behavior and transconductance. The calculations are calibrated using experimental data. The threshold voltage can be tuned from enhancement to depletion mode with wire doping. Surface states cause a shift of threshold voltage and saturation current. The saturation current depends on the gate design, with a composite gate acting as field plate in the p-body device. He joined Bell Laboratories, Murray Hill, NJ, as a Technical Staff Member. In October 2001, he joined the Optical Access and Transport Division, Agere Systems, Alhambra, CA. In 2004, he was appointed an Assistant Professor at ETH Zurich,. Since 2008, at the University of Kassel, Kassel, Germany, and he has been a Professor the Head of the Computational Electronics and Photonics Group, and co-director of CINSaT since 2010. His research interests include computational optoelectronics, process and device design of semiconductor photonic devices, microwave components, and electromagnetics modeling for nanophotonics. Dr. Witzigmann is a senior member of the SPIE and IEEE.

Free vibration analysis of single-walled boron nitride nanotubes based on a computational mechanics framework

Science.gov (United States)

Yan, J. W.; Tong, L. H.; Xiang, Ping

2017-12-01

Free vibration behaviors of single-walled boron nitride nanotubes are investigated using a computational mechanics approach. Tersoff-Brenner potential is used to reflect atomic interaction between boron and nitrogen atoms. The higher-order Cauchy-Born rule is employed to establish the constitutive relationship for single-walled boron nitride nanotubes on the basis of higher-order gradient continuum theory. It bridges the gaps between the nanoscale lattice structures with a continuum body. A mesh-free modeling framework is constructed, using the moving Kriging interpolation which automatically satisfies the higher-order continuity, to implement numerical simulation in order to match the higher-order constitutive model. In comparison with conventional atomistic simulation methods, the established atomistic-continuum multi-scale approach possesses advantages in tackling atomic structures with high-accuracy and high-efficiency. Free vibration characteristics of single-walled boron nitride nanotubes with different boundary conditions, tube chiralities, lengths and radii are examined in case studies. In this research, it is pointed out that a critical radius exists for the evaluation of fundamental vibration frequencies of boron nitride nanotubes; opposite trends can be observed prior to and beyond the critical radius. Simulation results are presented and discussed.

M5TM alloy high burnup behavior and worldwide licensing

International Nuclear Information System (INIS)

Mardon, J.P.; Hoffmann, P.B.; Garner, G.L.

2005-01-01

The in-reactor behavior of advanced PWR Zirconium alloys at burnups equal to or below licensing limits has been widely reported. Specifically, the advanced alloy M5 has demonstrated impressive improvements over Zircaloy-4 for fuel rod cladding and fuel assembly structural components. To demonstrate superiority of the alloy at burnups beyond current licensing limits, M5 has been operated in PWR at burnups exceeding 71 GWd/tU in the United States and 78 GWd/tU in Europe. Two extensive irradiation programs have been performed in the United States to demonstrate alloy M5 performance beyond current licensing limits. Four M5 TM fuel rods were exposed to four 24-month cycles in a 15x15 reactor beginning in 1995. Additionally, one 17x17 lead assembly containing M5 fuel rods and guide tubes was operated for four 18-month cycles beginning from 1997. Post-irradiation examinations (PIE) performed after all four cycles in the 15x15 demonstration program revealed excellent performance in the licensed burnup and in the high burnup stages of the experience. Examination of the 4th cycle 17x17 assembly will be accomplished in two stages the first of which is scheduled for June 2005. Moreover, several irradiation campaigns have been performed in Europe in order to confirm the excellent M5 in-pile behavior in demanding PWRs irradiation conditions with regard to void fraction, heat flux, lithium content and temperature. Results from the high burnup fuel examinations verify that the excellent performance achieved up to 62 GWd/tU was continued into higher burnup. The results of high burnup PIE campaigns for European and American PWR's are presented in this paper. Measured performance indicators include fuel assembly dimensional stability parameters (assembly length, fuel rod length, assembly bow, fuel rod bow, fuel rod radial creep and spacer grid width), oxidation measurements (fuel rod and guide tube) and hydrogen pick-up data (fuel rod). In the framework of PCI studies, power ramp

Effect of core burnup on the dynamic behavior of fast reactors

International Nuclear Information System (INIS)

Ilberg, D.; Saphier, D.; Yiftah, S.

1977-01-01

Performance of a dynamic analysis, taking burnup changes into account, requires fission-product nuclear data of relatively small uncertainty, suitable burnup calculation models, and dynamic computer programs. These were prepared and used with the following results: (1) Significant changes in static and dynamic parameters were observed when investigating the effect of burnup. These changes were found to be larger than differences introduced by the uncertainty of the fission-product nuclear data. (2) A one-dimensional burnup computer program was prepared. It was found that a burnup model based on the generalized radioactive decay scheme is suitable for accurate fast reactor calculations. (3) Space-time dynamic calculations of fast reactors having different burnup levels were performed. The stability difference between ''clean'' and high burnup cores is greater when local rather than uniform perturbations are inserted along the entire core length. The magnitude by which the ''end-of-life'' core increases the transient excursion over that of the clean core depends on the particular region in which the perturbation is inserted. The end-of-life core will magnify the transient excursion more than the clean core whenever the perturbation is inserted into a region having a higher adjoint flux level than that of the clean core. However, when a reactor safety system operates successfully, the difference in the temperature transient of the clean and end-of-life cores will be relatively small. It is suggested that only the analysis of large local perturbations be performed for end-of-life cores as well as for clean cores in the safety evaluation of fast reactors

Comparison of analysis methods for burnup credit applications

International Nuclear Information System (INIS)

Sanders, T.L.; Brady, M.C.; Renier, J.P.; Parks, C.V.

1989-01-01

The current approach used for the development and certification of spent fuel storage and transport casks requires an assumption of fresh fuel isotopics in the criticality safety analysis. However, it has been shown that there is a considerable reactivity reduction when the isotopics representative of the depleted (or burned) fuel are used in a criticality analysis. Thus, by taking credit for the burned state of the fuel (i.e., burnup credit), a cask designer could achieve a significant increase in payload. Accurate prediction of k eff for spent fuel arrays depends both on the criticality safety analysis and the prediction of the spent fuel isotopics via a depletion analysis. Spent fuel isotopics can be obtained from detailed multidimensional reactor analyses, e.g. the code PDQ, or from point reactor burnup models. These reactor calculations will help verify the adequacy of the isotopics and determine Δk eff biases for various analysis assumptions (with and without fission products, actinide absorbers, burnable poison rods, etc.). New software developed to interface PDQ multidimensional isotopics with KENO V.a reactor and cask models is described. Analyses similar to those performed for the reactor cases are carried out with a representative burnup credit cask model using the North Anna fuel. This paper presents the analysis methodology that has been developed for evaluating the physics issues associated with burnup credit. It is applicable in the validation and characterization of fuel isotopics as well as in determining the influence of various analysis assumptions in terms of δk eff . The methodology is used in the calculation of reactor restart criticals and analysis of a typical burnup credit cask

Interlayer shear of nanomaterials: Graphene-graphene, boron nitride-boron nitride and graphene-boron nitride

Institute of Scientific and Technical Information of China (English)

Yinfeng Li; Weiwei Zhang; Bill Guo; Dibakar Datta

2017-01-01

In this paper,the interlayer sliding between graphene and boron nitride (h-BN) is studied by molecular dynamics simulations.The interlayer shear force between h-BN/h-BN is found to be six times higher than that of graphene/graphene,while the interlayer shear between graphene/h-BN is approximate to that of graphene/graphene.The graphene/h-BN heterostructure shows several anomalous interlayer shear characteristics compared to its bilayer counterparts.For graphene/graphene and h-BN/h-BN,interlayer shears only exit along the sliding direction while interlayer shear for graphene/h-BN is observed along both the translocation and perpendicular directions.Our results provide significant insight into the interlayer shear characteristics of 2D nanomaterials.

The octopus burnup and criticality code system

Energy Technology Data Exchange (ETDEWEB)

Kloosterman, J.L.; Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de

1996-09-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

The OCTOPUS burnup and criticality code system

Energy Technology Data Exchange (ETDEWEB)

Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

1996-06-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.).

The octopus burnup and criticality code system

International Nuclear Information System (INIS)

Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.

1996-01-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

The OCTOPUS burnup and criticality code system

International Nuclear Information System (INIS)

Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de

1996-06-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)

Nitride stabilized core/shell nanoparticles

Science.gov (United States)

Kuttiyiel, Kurian Abraham; Sasaki, Kotaro; Adzic, Radoslav R.

2018-01-30

Nitride stabilized metal nanoparticles and methods for their manufacture are disclosed. In one embodiment the metal nanoparticles have a continuous and nonporous noble metal shell with a nitride-stabilized non-noble metal core. The nitride-stabilized core provides a stabilizing effect under high oxidizing conditions suppressing the noble metal dissolution during potential cycling. The nitride stabilized nanoparticles may be fabricated by a process in which a core is coated with a shell layer that encapsulates the entire core. Introduction of nitrogen into the core by annealing produces metal nitride(s) that are less susceptible to dissolution during potential cycling under high oxidizing conditions.

Origin of interfacial charging in irradiated silicon nitride capacitors

International Nuclear Information System (INIS)

Hughes, R.C.

1984-01-01

Many experiments show that when metal-silicon nitride-silicon dioxide-silicon (MNOS) devices are irradiated in short circuit, a large interfacial charge builds up near the nitride-SiO 2 -Si interface. This effect cannot be explained by simple models of radiation-induced conductivity of the nitride, but it is reported here that inclusion of carrier diffusion and recombination in the photoconductivity equations can predict the observed behavior. Numerical solutions on a computer are required, however, when these complications are added. The simulations account for the magnitude and radiation dose dependence of the results, as well as the occurrence of a steady state during the irradiation. The location of the excess trapped charge near the interface is also predicted, along with the large number of new traps which must be introduced to influence the steady-state charge distribution

Sophistication of burnup analysis system for fast reactor (2)

International Nuclear Information System (INIS)

Yokoyama, Kenji; Hirai, Yasushi; Tatsumi, Masahiro

2010-10-01

Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

Technical Development on Burn-up Credit for Spent LWR Fuel

Energy Technology Data Exchange (ETDEWEB)

Gauld, I.C.

2001-12-26

Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

Technical development on burn-up credit for spent LWR fuels

International Nuclear Information System (INIS)

Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori

2000-10-01

Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

Technical development on burn-up credit for spent LWR fuels

Energy Technology Data Exchange (ETDEWEB)

Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-10-01

Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

Directory of Open Access Journals (Sweden)

Gajda Paweł

2016-01-01

Full Text Available Monte Carlo methodology provides reference statistical solution of neutron transport criticality problems of nuclear systems. Estimated reaction rates can be applied as an input to Bateman equations that govern isotopic evolution of reactor materials. Because statistical solution of Boltzmann equation is computationally expensive, it is in practice applied to time steps of limited length. In this paper we show that simple staircase step model leads to underprediction of numerical fuel burnup (Fissions per Initial Metal Atom – FIMA. Theoretical considerations indicates that this error is inversely proportional to the length of the time step and origins from the variation of heating per source neutron. The bias can be diminished by application of predictor-corrector step model. A set of burnup simulations with various step length and coupling schemes has been performed. SERPENT code version 1.17 has been applied to the model of a typical fuel assembly from Pressurized Water Reactor. In reference case FIMA reaches 6.24% that is equivalent to about 60 GWD/tHM of industrial burnup. The discrepancies up to 1% have been observed depending on time step model and theoretical predictions are consistent with numerical results. Conclusions presented in this paper are important for research and development concerning nuclear fuel cycle also in the context of Gen4 systems.

Boron nitride encapsulated graphene infrared emitters

International Nuclear Information System (INIS)

Barnard, H. R.; Zossimova, E.; Mahlmeister, N. H.; Lawton, L. M.; Luxmoore, I. J.; Nash, G. R.

2016-01-01

The spatial and spectral characteristics of mid-infrared thermal emission from devices containing a large area multilayer graphene layer, encapsulated using hexagonal boron nitride, have been investigated. The devices were run continuously in air for over 1000 h, with the emission spectrum covering the absorption bands of many important gases. An approximate solution to the heat equation was used to simulate the measured emission profile across the devices yielding an estimated value of the characteristic length, which defines the exponential rise/fall of the temperature profile across the device, of 40 μm. This is much larger than values obtained in smaller exfoliated graphene devices and reflects the device geometry, and the increase in lateral heat conduction within the devices due to the multilayer graphene and boron nitride layers.

Boron nitride encapsulated graphene infrared emitters

Energy Technology Data Exchange (ETDEWEB)

Barnard, H. R.; Zossimova, E.; Mahlmeister, N. H.; Lawton, L. M.; Luxmoore, I. J.; Nash, G. R., E-mail: g.r.nash@exeter.ac.uk [College of Engineering, Mathematics and Physical Sciences, University of Exeter, Exeter EX4 4QF (United Kingdom)

2016-03-28

The spatial and spectral characteristics of mid-infrared thermal emission from devices containing a large area multilayer graphene layer, encapsulated using hexagonal boron nitride, have been investigated. The devices were run continuously in air for over 1000 h, with the emission spectrum covering the absorption bands of many important gases. An approximate solution to the heat equation was used to simulate the measured emission profile across the devices yielding an estimated value of the characteristic length, which defines the exponential rise/fall of the temperature profile across the device, of 40 μm. This is much larger than values obtained in smaller exfoliated graphene devices and reflects the device geometry, and the increase in lateral heat conduction within the devices due to the multilayer graphene and boron nitride layers.

Fabrication of vanadium nitride by carbothermal nitridation reaction

International Nuclear Information System (INIS)

Wang Xitang; Wang Zhuofu; Zhang Baoguo; Deng Chengji

2005-01-01

Vanadium nitride is produced from V 2 O 5 by carbon-thermal reduction and nitridation. When the sintered temperature is above 1273 K, VN can be formed, and the nitrogen content of the products increased with the firing temperature raised, and then is the largest when the sintered temperature is 1573 K. The C/V 2 O 5 mass ratio of the green samples is the other key factor affecting on the nitrogen contents of the products. The nitrogen content of the products reaches the most when the C/V 2 O 5 mass ratio is 0.33, which is the theoretical ratio of the carbothermal nitridation of V 2 O 5 . (orig.)

Simulated radiation effects in the superinsulating phase of titanium nitride films

Directory of Open Access Journals (Sweden)

Vujisić Miloš Lj.

2011-01-01

Full Text Available This paper investigates possible effects of alpha particle and ion beam irradiation on the properties of the superinsulating phase, recently observed in titanium nitride films, by using numerical simulation of particle transport. Unique physical properties of the superinsulating state are considered by relying on a two-dimensional Josephson junction array as a model of material structure. It is suggested that radiation-induced change of the Josephson junction charging energy would not affect the current-voltage characteristics of the superinsulating film significantly. However, it is theorized that a relapse to an insulating state with thermally activated resistance is possible, due to radiation-induced disruption of the fine-tuned granular structure. The breaking of Cooper pairs caused by incident and displaced ions may also destroy the conditions for a superinsulating phase to exist. Finally, even the energy loss to phonons can influence the superinsulating state, by increasing the effective temperature of the phonon thermostat, thereby reestablishing means for an energy exchange that can support Cooper pair tunneling.

Molecular dynamics simulation of nano-indentation of (111) cubic boron nitride with optimized Tersoff potential

International Nuclear Information System (INIS)

Zhao, Yinbo; Peng, Xianghe; Fu, Tao; Huang, Cheng; Feng, Chao; Yin, Deqiang; Wang, Zhongchang

2016-01-01

Highlights: • We optimize Tersoff potential to simulate the cBN better under nanoidentation. • Dislocations slip more easily along and directions on the {111} plane. • Shuffle-set dislocation slip along direction on {111} plane first. • A tetrahedron structure is found in the initial stage of the indentation. - Abstract: We conduct molecular dynamics simulation of nanoindentation on (111) surface of cubic boron nitride and find that shuffle-set dislocations slip along direction on {111} plane at the initial stage of the indentation. The shuffle-set dislocations are then found to meet together, forming surfaces of a tetrahedron. We also find that the surfaces are stacking-fault zones, which intersect with each other, forming edges of stair-rod dislocations along direction. Moreover, we also calculate the generalized stacking fault (GSF) energies along various gliding directions on several planes and find that the GSF energies of the {111} and {111} systems are relatively smaller, indicating that dislocations slip more easily along and directions on the {111} plane.

Discharge Burnup Evaluation of Natural Uranium Loaded CANFLEX-43 Fuel Bundle

International Nuclear Information System (INIS)

Roh, Gyu Hong; Kim, Yong Hee; Kim, Won Young; Park, Joo Hwan

2009-11-01

Using WIMS-AECL code, which is 2-dimensional lattice core used in CANDU physics calculation, the discharge burnup of the natural uranium loaded CANFLEX-43 fuel bundle was evaluated by comparing the discharge burnup of standard 37 element fuel bundle. When the discharge burnup of the standard 37 element fuel is 7,200 MWd/MTU, that of the CANFLEX 43 fuel bundle was evaluated as 7,077 MWd/MTU, by applying the same lattice conditions for both fuel bundles

Status of burnup credit implementation and research in Switzerland

International Nuclear Information System (INIS)

Grimm, P.

2001-01-01

Burnup credit has recently been approved by the Swiss licensing authority for the spent-fuel storage pool of a PWR plant for fuel exceeding the originally licensed initial enrichment. The criticality safety assessment is based on a configuration consisting of a small number (approximately a reload batch) of fresh assemblies surrounded by assemblies having a burnup corresponding to the minimum value in the top 1 m section after one cycle of irradiation. The allowable initial enrichment in this configuration is about 0.5% higher than for all fresh fuel. A central storage facility for all types of radioactive wastes from Switzerland, including cask storage of spent fuel assemblies is being commissioned presently. The first applications for licenses for casks to be used in this facility have been submitted. Credit for burnup has not been requested in these applications (conforming to the original licenses of the casks in their countries of origin), but utilities are interested in burnup credit for fuel with higher initial enrichments. Reactivity worth measurements as well as chemical assays of spent fuel samples in the LWR-PROTEUS facility at PSI are in detailed planning currently. The experiments, scheduled to start in 2001, will be performed in cooperation with the Swiss utilities and their fuel vendors. Although the focus of interest of these partners is on validation of in-core fuel management tools, the same experiments are also applicable to burnup credit, and contacts with further potential partners interested in this field are underway. (author)

A guide to introducing burnup credit, preliminary version (English translation)

International Nuclear Information System (INIS)

Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu

2017-06-01

There is an ongoing discussion on the application of burnup credit to the criticality safety controls of facilities that treat spent fuels. With regard to such application of burnup credit in Japan, this document summarizes the current technical status of the prediction of the isotopic composition and criticality of spent fuels, as well as safety evaluation concerns and the current status of legal affairs. This report is an English translation of A Guide to Introducing Burnup Credit, Preliminary Version, originally published in Japanese as JAERI-Tech 2001-055 by the Nuclear Fuel Cycle Facility Safety Research Committee. (author)

BNFL assessment of methods of attaining high burnup MOX fuel

International Nuclear Information System (INIS)

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

1998-01-01

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

Approach for implementing burnup credit in high-capacity truck casks

International Nuclear Information System (INIS)

Boshoven, J.; Hopf, J.; Su, S.

1991-01-01

General Atomics (GA) will be submitting an application for certification to the US Nuclear Regulatory Commission (NRC) for the GA-4 and GA-9 Casks in 1992. To maintain a capacity of four pressurized-water-reactor (PWR) spent fuel assemblies, the GA-4 Cask uses burnup credit as part of the criticality control for the higher enrichments. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA presents here an approach to burnup credit analysis to be included in the Safety Analysis Report for Packaging (SARP). 6 refs., 2 figs., 5 tabs

Measurement and interpretation of triton burnup in Jet deuterium plasmas

International Nuclear Information System (INIS)

Jarvis, O.N.; Kallne, J.; Sadler, G.; van Belle, P.; Gorini, G.; Conroy, S.; Verschuur, K.

1989-01-01

The confinement and slowing down of fast tritons in JET deuterium plasmas is investigated. The ratio of 14 MeV and 2.5 MeV neutron production rates is measured. This ratio is equal to the fraction of tritons which burnup. The 2.5 MeV neutron emission is obtained from a set of fission chambers for which the calibration uncertainty is about 10%. The absolute calibration of the activation technique is calculated. The comparison between experimental and theoretical burnup ratios, for JET 1987 data, is shown. The range of conditions over which measurements of triton burnup fraction were obtained, is illustrated

Advances in Metallic Fuels for High Burnup and Actinide Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.

2016-10-01

Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.

Method for producing polycrystalline boron nitride

International Nuclear Information System (INIS)

Alexeevskii, V.P.; Bochko, A.V.; Dzhamarov, S.S.; Karpinos, D.M.; Karyuk, G.G.; Kolomiets, I.P.; Kurdyumov, A.V.; Pivovarov, M.S.; Frantsevich, I.N.; Yarosh, V.V.

1975-01-01

A mixture containing less than 50 percent of graphite-like boron nitride treated by a shock wave and highly defective wurtzite-like boron nitride obtained by a shock-wave method is compressed and heated at pressure and temperature values corresponding to the region of the phase diagram for boron nitride defined by the graphite-like compact modifications of boron nitride equilibrium line and the cubic wurtzite-like boron nitride equilibrium line. The resulting crystals of boron nitride exhibit a structure of wurtzite-like boron nitride or of both wurtzite-like and cubic boron nitride. The resulting material exhibits higher plasticity as compared with polycrystalline cubic boron nitride. Tools made of this compact polycrystalline material have a longer service life under impact loads in machining hardened steel and chilled iron. (U.S.)

COMRAD96, Nuclear Fuel Burnup and Depletion Calculation System

International Nuclear Information System (INIS)

Suyama, K.; Masukawa, F.; Ido, M.; Enomoto, M.; Takyu, S.; Hara, T.

2002-01-01

1 - Description of program or function: Burn-up calculation of nuclear fuel. 2 - Methods: Matrix exponential method, Bateman Equation. 3 - Restrictions on the complexity of the problem: a) One-grouped cross section library should be prepared for the fuel system to be analyzed using UNITBURN. However, UNITBURN is not available now for UNIX systems. b) Gamma ray spectrometry calculation will fail using the attached piflib routine. This problem has already been rectified in the internal version. 4 - Typical running time: Two minutes for standard burn-up calculation on Sun ULTRA 30. 5 - Unusual features - a) Selection of Matrix exponential method, or Bateman Equation. b) JDDL, a detailed decay chain data based on ENSDF. 6 - Related or auxiliary programs: UNITBURN: Burnup calculation code unit cell system

The use of burnup credit for spent fuel cask design

International Nuclear Information System (INIS)

Lake, W.H.

1993-01-01

A new generation of high capacity spent fuel transport casks is being developed by the U.S. Department of Energy (DOE) as part of the Federal Waste Management System (FWMS). Burnup credit, which recognizes the reduced reactivity of spent fuel is being used for these casks. Two cask designs being developed for DOE by Babcock and Wilcox and General Atomics use burnup credit. The cask designs must be certified by the U.S. Nuclear Regulatory Commission (NRC) if they are to be used in the FWMS. Certification of these casks by the NRC would not require any change in the NRC's transport regulations, and would be consistent with past practices. Furthermore, use of burnup credit casks appears to be consistent with current International Atomic Energy Agency (IAEA) rules and regulations. To support NRC certification, DOE has identified the technical issues related to burnup credit, and embarked on a development program to resolve them. (J.P.N.)

Effect of core configuration on the burnup calculations of MTR research reactors

International Nuclear Information System (INIS)

Hussein, H.M.; Amin, E.H.; Sakr, A.M.

2014-01-01

Highlights: • 3D burn-up calculations of MTR-type research reactor were performed. Examination of the effect of control rod pattern on power density and neutron flux distributions is presented. • The calculations are performed using the MTR P C package and the programs (WIMS and CITVAP). • An empirical formula was generated for every fuel element type, to correlate irradiation to burn-up. - Abstract: In the present paper, three-dimensional burn-up calculations were performed using different patterns of control rods, in order to examine their effect on power density and neutron flux distributions through out the entire core and hence on the local burn-up distribution. These different cores burn-up calculations are carried out for an operating cycle equivalent to 15 Full Power Days (FPDs), with a power rating of 22 MW. Calculations were performed using an example of a typical research reactor of MTR-type using the internationally known computer codes’ package “MTR P C system”, using the cell calculation transport code WIMS-D4 with 12 energy groups and the core calculation diffusion code CITVAP with 5 energy groups. A depletion study was done and the effects on the research reactor fuel (U-235) were performed. The burn-up percentage (B.U.%) curves for every fuel element type were drawn versus irradiation (MWD/TE). Then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Charts of power density and neutron flux distribution for each core were plotted at different sections of each fuel element of the reactor core. Then a complete discussion and analysis of these curves are performed with comparison between the different core configurations, illustrating the effect of insertion or extraction of either of the four control rods directly on the neutron flux and consequently on the power distribution and burn-up. A detailed study of fuel burn-up gives detailed insight on the different B.U.% calculations

Technical Issues in the development of high burnup and long cycle fuel pellets

Energy Technology Data Exchange (ETDEWEB)

Kim, Dong Joo; Yang, Jae Ho; Oh, Jang Soo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Nam, Ik Hui [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

Over the last half century, a nuclear fuel cycle, a fuel discharged burnup and a uranium enrichment of the LWR (Light Water Reactor) fuel have continuously increased. It was the efforts to reduce the LWR fuel cycle cost, and to make reactor operation more efficiently. Improved fuel and reactor performance contribute further to the reduction and management efficiency of spent fuels. The primary incentive for operating nuclear reactor fuel to higher burnup and longer cycle is the economic benefits. The fuel cycle costs could be reduced by extending fuel discharged burnup and fuel cycle length. The higher discharged burnup can increase the energy production per unit fuel mass or fuel assembly. The longer fuel cycle can increase reactor operation flexibility and reduce the fuel changing operation and the spent fuel management burden. The margin to storage capacity limits would be also increased because high burnup and long cycle fuel reduces the mass of spent fuels. However, increment of fuel burnup and cycle length might result in the acceleration of material aging consisting fuel assembly. Then, the safety and integrity of nuclear fuel will be degraded. Therefore, to simultaneously enhance the safety and economics of the LWR fuel through the fuel burnup and cycle extension, it is indispensable to develop the innovative nuclear fuel material concepts and technologies which can overcome degradation of fuel safety. New fuel research project to extend fuel discharged burnup and cycle length has been launched in KAERI. Main subject is to develop innovative LWR fuel pellets which can provide required fuel performance and safety at extended fuel burnup and cycle length. In order to achieve the mission, we need to know that what the impediments are and how to break through current limit of fuel pellet properties. In this study, the technical issues related to fuel pellets at high burnup were surveyed and summarized. We have collected the technical issues in the literatures

Technical Issues in the development of high burnup and long cycle fuel pellets

International Nuclear Information System (INIS)

Kim, Dong Joo; Yang, Jae Ho; Oh, Jang Soo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Nam, Ik Hui

2012-01-01

Over the last half century, a nuclear fuel cycle, a fuel discharged burnup and a uranium enrichment of the LWR (Light Water Reactor) fuel have continuously increased. It was the efforts to reduce the LWR fuel cycle cost, and to make reactor operation more efficiently. Improved fuel and reactor performance contribute further to the reduction and management efficiency of spent fuels. The primary incentive for operating nuclear reactor fuel to higher burnup and longer cycle is the economic benefits. The fuel cycle costs could be reduced by extending fuel discharged burnup and fuel cycle length. The higher discharged burnup can increase the energy production per unit fuel mass or fuel assembly. The longer fuel cycle can increase reactor operation flexibility and reduce the fuel changing operation and the spent fuel management burden. The margin to storage capacity limits would be also increased because high burnup and long cycle fuel reduces the mass of spent fuels. However, increment of fuel burnup and cycle length might result in the acceleration of material aging consisting fuel assembly. Then, the safety and integrity of nuclear fuel will be degraded. Therefore, to simultaneously enhance the safety and economics of the LWR fuel through the fuel burnup and cycle extension, it is indispensable to develop the innovative nuclear fuel material concepts and technologies which can overcome degradation of fuel safety. New fuel research project to extend fuel discharged burnup and cycle length has been launched in KAERI. Main subject is to develop innovative LWR fuel pellets which can provide required fuel performance and safety at extended fuel burnup and cycle length. In order to achieve the mission, we need to know that what the impediments are and how to break through current limit of fuel pellet properties. In this study, the technical issues related to fuel pellets at high burnup were surveyed and summarized. We have collected the technical issues in the literatures

Preparation of uranium nitride

International Nuclear Information System (INIS)

Potter, R.A.; Tennery, V.J.

1976-01-01

A process is described for preparing actinide-nitrides from massive actinide metal which is suitable for sintering into low density fuel shapes by partially hydriding the massive metal and simultaneously dehydriding and nitriding the dehydrided portion. The process is repeated until all of the massive metal is converted to a nitride

Determination of nuclear fuel burn-up using mass spectrometric techniques

International Nuclear Information System (INIS)

Saha, B.; Bagyalakshmi, R.; Periaswami, G.; Kavimandan, V.D.; Chitambar, S.A.; Jain, H.C.; Mathews, C.K.

1977-01-01

Determination of burn-up using a stable fission product monitor such as 148 Nd and heavy elements, determined by isotope dilution mass spectrometry gives the most accurate data. This report describes the work carried out to standardise the conditions for burn-up determination. Some typical results are given. (author)

Burnup credit feasibility for BWR spent fuel shipments

International Nuclear Information System (INIS)

Broadhead, B.L.

1990-01-01

Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

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)

Restructuring of burnup sensitivity analysis code system by using an object-oriented design approach

International Nuclear Information System (INIS)

Kenji, Yokoyama; Makoto, Ishikawa; Masahiro, Tatsumi; Hideaki, Hyoudou

2005-01-01

A new burnup sensitivity analysis code system was developed with help from the object-oriented technique and written in Python language. It was confirmed that they are powerful to support complex numerical calculation procedure such as reactor burnup sensitivity analysis. The new burnup sensitivity analysis code system PSAGEP was restructured from a complicated old code system and reborn as a user-friendly code system which can calculate the sensitivity coefficients of the nuclear characteristics considering multicycle burnup effect based on the generalized perturbation theory (GPT). A new encapsulation framework for conventional codes written in Fortran was developed. This framework supported to restructure the software architecture of the old code system by hiding implementation details and allowed users of the new code system to easily calculate the burnup sensitivity coefficients. The framework can be applied to the other development projects since it is carefully designed to be independent from PSAGEP. Numerical results of the burnup sensitivity coefficient of a typical fast breeder reactor were given with components based on GPT and the multicycle burnup effects on the sensitivity coefficient were discussed. (authors)

SRAC-95, Cell Calculation with Burnup, Fuel Management for Thermal Reactors

International Nuclear Information System (INIS)

Tsuchihashi, K.; Ishiguro, Y.; Kaneko, K.; Ido, M.

2004-01-01

1 - Description of program or function: General neutronics calculation including cell calculation with burn-up, core calculation for any type of thermal reactor. Core burn-up calculation and fuel management by an auxiliary code. 2 - Method of solution: Collision probability method, 1D and 2D Sn for cell calculation; 1D, 2D and 3D diffusion for core calculation. 3 - Restrictions on the complexity of the problem: 20 regions for a continuous energy resonance absorption calculation and 16 steps for cell burn-up

Modelling of phenomena associated with high burnup fuel behaviour during overpower transients

International Nuclear Information System (INIS)

Sills, H.E.; Langman, V.J.; Iglesias, F.C.

1995-01-01

Phenomena of importance to the behaviour of high burnup fuel subjected to conditions of rapid overpower (i.e., LWR RIAs) include the change in cladding material properties due to irradiation, pellet-clad interaction (PCI) and 'rim' effects associated with the periphery of high burnup fuel. 'Rim' effects are postulated to be caused by changes in fuel morphology at high burnup. Typical discharge burnups for CANDU fuel are low compared to LWRs. Maximum linear ratings for CANDU fuel are higher than those for LWRs. However, under normal operating conditions, the Zircaloy-4 clad of the CANDU fuel is collapsed onto the fuel stack. Thus, the CANDU fuel performance codes model the transient behaviour of the fuel-to-clad interface and are capable of assessing the potential for pellet-clad mechanical interaction (PCMI) failures for a wide range of overpower conditions. This report provides a discussion of the modelling of the phenomena of importance to high burnup fuel behaviour during rapid overpower transients. (author)

A survey of previous and current industry-wide efforts regarding burnup credit

International Nuclear Information System (INIS)

Jones, R.H.

1989-01-01

Sandia has examined the matter of burnup credit from the perspective of physics, logistics, risk, and economics. A limited survey of the nuclear industry has been conducted to get a feeling for the actual application of burnup credit. Based on this survey, it can be concluded that the suppliers of spent fuel storage and transport casks are in general agreement that burnup credit offers the potential for improvements in cask efficiency without increasing the risk of accidental criticality. The actual improvement is design-specific but limited applications have demonstrated that capacity increases in the neighborhood of 20 percent are not unrealistic. A number of these vendors acknowledge that burnup credit has not been reduced to practice in cask applications and suggest that operational considerations may be more important to regulatory acceptance than to the physics. Nevertheless, the importance of burnup credit to the nuclear industry as a cask design and analysis tool has been confirmed by this survey

Post-irradiation examinations of inert matrix nitride fuel irradiated in JMTR (01F-51A capsule)

International Nuclear Information System (INIS)

Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Honda, Junichi; Hatakeyama, Yuichi; Ono, Katsuto; Matsui, Hiroki; Arai, Yasuo

2007-03-01

A plutonium nitride fuel pin containing inert matrix such as ZrN and TiN was encapsulated in 01F-51A and irradiated in JMTR. Minor actinides are surrogated by plutonium. Average linear powers and burnups were 408W/cm, 30000MWd/t(Zr+Pu) [132000MWd/t-Pu] for (Zr,Pu)N and 355W/cm, 38000MWd/t(Ti+Pu) [153000MWd/t-Pu] for (TiN,PuN). The irradiated capsule was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pin. Very low fission gas release rate of about 1.6% was measured. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

Calculation of pellet radial power distributions with a Monte Carlo burnup code

International Nuclear Information System (INIS)

Suzuki, Motomu; Yamamoto, Toru; Nakata, Tetsuo

2010-01-01

The Japan Nuclear Energy Safety Organization (JNES) has been working on an irradiation test program of high-burnup MOX fuel at Halden Boiling Water Reactor (HBWR). MOX and UO 2 fuel rods had been irradiated up to about 64 GWd/t (rod avg.) as a Japanese utilities research program (1st phase), and using those fuel rods, in-situ measurement of fuel pellet centerline temperature was done during the 2nd phase of irradiation as the JNES test program. As part of analysis of the temperature data, power distributions in a pellet radial direction were analyzed by using a Monte Carlo burnup code MVP-BURN. In addition, the calculated results of deterministic burnup codes SRAC and PLUTON for the same problem were compared with those of MVP-BURN to evaluate their accuracy. Burnup calculations with an assembly model were performed by using MVP-BURN and those with a pin cell model by using SRAC and PLUTON. The cell pitch and, therefore, fuel to moderator ratio in the pin cell calculation was determined from the comparison of neutron energy spectra with those of MVP-BURN. The fuel pellet radial distributions of burnup and fission reaction rates at the end of the 1st phase irradiation were compared between the three codes. The MVP-BURN calculation results show a large peaking in the burnup and fission rates in the pellet outer region for the UO 2 and MOX pellets. The SRAC calculations give very close results to those of the MVP-BURN. On the other hand, the PLUTON calculations show larger burnup for the UO 2 and lower burnup for the MOX pellets in the pellet outer region than those of MVP-BURN, which lead to larger fission rates for the UO 2 and lower fission rates for the MOX pellets, respectively. (author)

Burnup calculation methodology in the serpent 2 Monte Carlo code

International Nuclear Information System (INIS)

Leppaenen, J.; Isotalo, A.

2012-01-01

This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)

Process for the production of metal nitride sintered bodies and resultant silicon nitride and aluminum nitride sintered bodies

Science.gov (United States)

Yajima, S.; Omori, M.; Hayashi, J.; Kayano, H.; Hamano, M.

1983-01-01

A process for the manufacture of metal nitride sintered bodies, in particular, a process in which a mixture of metal nitrite powders is shaped and heated together with a binding agent is described. Of the metal nitrides Si3N4 and AIN were used especially frequently because of their excellent properties at high temperatures. The goal is to produce a process for metal nitride sintered bodies with high strength, high corrosion resistance, thermal shock resistance, thermal shock resistance, and avoidance of previously known faults.

Simulation of STM technique for electron transport through boron-nitride nanotubes

International Nuclear Information System (INIS)

Ganji, M.D.; Mohammadi-nejad, A.

2008-01-01

We report first-principles calculations on the electrical transport properties of boron-nitrid nanotubes (BNNTs). We consider a single walled (5,0) boron-nitrid nanotube sandwiched between an Au(1 0 0) substrate and a monatomic Au scanning tunneling microscope (STM) tip. Lateral motion of the tip over the nanotube wall cause it to change from one conformation class to the others and to switch between a strongly and a weakly conducting state. Thus, surprisingly, despite their apparent simplicity these Au/BNNT/Au nanowires are shown to be a convenient switch. Experiments with a conventional STM are proposed to test these predictions. The projection of the density of states (PDOS) and the transmission coefficients T(E) of the two-probe systems at zero bias are analyzed, and it suggests that the variation of the coupling between the wire and the electrodes leads to switching behaviour

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Role of measurement systems in burnup credit operations

International Nuclear Information System (INIS)

Ewing, R.I.; Sanders, T.L.

1991-01-01

Spent fuel transport casks designed using burnup credit have increased payloads that may greatly reduce the number of shipments required to transport spent fuel from reactor sites to repositories. Burnup credit is obtained by applying the reduced reactivity of spent fuel to considerations of nuclear criticality in the design of transport casks. Although it does not appear to be possible to directly measure the criticality of spent fuel assemblies, measurements can be employed to ensure that the only assemblies loaded into a cask have the characteristics appropriate to that cask design. An effective on-site measurement system must be matched to the characteristics of the spent fuel cask design and to the inventory of spent fuel. For operation reasons the system should be simple, accurate, efficient, and easily calibrated. This paper is part of a study to examine the effects of the spent fuel inventory in the U.S. on the selection of measurement systems useful in burnup credit operations

Plasmonic Titanium Nitride Nanostructures via Nitridation of Nanopatterned Titanium Dioxide

DEFF Research Database (Denmark)

Guler, Urcan; Zemlyanov, Dmitry; Kim, Jongbum

2017-01-01

Plasmonic titanium nitride nanostructures are obtained via nitridation of titanium dioxide. Nanoparticles acquired a cubic shape with sharper edges following the rock-salt crystalline structure of TiN. Lattice constant of the resulting TiN nanoparticles matched well with the tabulated data. Energy...

CARMEN-SYSTEM, Programs System for Thermal Neutron Diffusion and Burnup with Feedback

International Nuclear Information System (INIS)

Ahnert, Carol; Aragones, Jose M.

1983-01-01

1 - Description of problem or function: CARMEN is a system of programs developed for the neutronic calculation of PWR cycles. It includes the whole chain of analysis from cell calculations to core calculations with burnup. The core calculations are based on diffusion theory with cross sections depending on the relevant space-dependent feedback effects which are present at each moment along the cycles. The diffusion calculations are in one, two or three dimensions and in two energy groups. The feedback effects which are treated locally are: burnup, water density, power density and fission products. In order to study in detail these parameters the core should be divided into as many zones as different cross section sets are expected to be required in order to reproduce reality correctly. A relevant difference in any feedback parameter between zones produces different cross section sets for the corresponding zones. CARMEN is also capable to perform the following calculations: - Multiplication factor by burnup step with fixed boron concentration - Buckling and control rod insertion - Buckling search by burnup step - Boron search by burnup step - Control rod insertion search by burnup step. 2 - Method of solution: The cell code (LEOPARD-TRACA) generates the fuel assembly cross sections versus burnup. This is the basic library to be used in the CARMEN code proper. With a planar distribution guess for power density, water density and fluxes, the macroscopic cross sections by zone are calculated by CARMEN, and then a diffusion calculation is done in the whole geometry. With the distribution of power density, heat accumulated in the coolant and the thermal and fast fluxes determined in the diffusion calculation, CARMEN calculates the values of the most relevant parameters that influence the macroscopic cross sections by zone: burnup, water density, effective fuel temperature and fission product concentrations. If these parameters by zone are different from the reference

The radial distribution of plutonium in high burnup UO2 fuels

International Nuclear Information System (INIS)

Lassmann, K.; O'Carroll, C.; Laar, J. van de; Walker, C.T.

1994-01-01

A new model (TUBRNP) is described which predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of uranium and plutonium isotopes. Comparisons between measurements and the predictions of the TUBRNP model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnups between 21 000 and 64 000 MWd/t. It is shown to be in excellent agreement with experimental measurements and is a marked improvement on earlier versions. (orig.)

Criterion for burn-up conditions in gas-cooled cryogenic current leads

International Nuclear Information System (INIS)

Bejan, A.; Cluss, E.M. Jr.

1976-01-01

Superconducting magnets are energized through helium vapour-cooled cryogenic current leads operating at high ratios of current to mass flow. The high current operation where lead temperature, runaway, and eventual burn-up are likely to occur is investigated. A simple criterion for estimating the burn-up operation conditions (current, mass flow) for a given lead geometry (cross-sectional area, length, heat exchanger area) is presented. This article stresses the role played by the available heat exchanger area in avoiding burn-up at high ratios of current to mass flow. (author)

The research on burnup characteristic of doping burnable poison in PWR

International Nuclear Information System (INIS)

Qiang Shenglong; Qin Dong; Chai Xiaoming; Yao Dong

2014-01-01

In PWR core design, burnable poisons are usually used for reactive compensation and power flatten. The choice of burnable poisons and how to match burnup would be the key-points for a long-life core design. We study the burnup character of doping burnable poisons (such as natural element, manual nuclide and soluble boron) in the PWR by the core burnup code MOI based on Monte Carlo method. The results show that Hf, Er and Eu doping burnable poison would be applicable for the nuclear design research on the long-life PWR core. (authors)

DELIGHT-B/REDEL, point reactivity burnup code for high-temperature gas-cooled reactor cells

International Nuclear Information System (INIS)

Shindo, Ryuiti; Watanabe, Takashi.

1977-03-01

Code DELIGHT-2 was previously developed to analyze cell burnup characteristics and to produce few-group constants for core burnup calculation in high-temperature gas-cooled reactors. In the code, burnup dependency of the burnable poison, boron-10, is considered with the homogeneous model of space. In actuality, however, the burnable poison is used as homogeneous rods or uniform rods of small granular poison and graphite, to control the reactivity and power distribution. Precise analysis of the burnup characteristics is thus difficult because of the heterogeneity due to the configuration of poison rods. In cell burnup calculation, the DELIGHT-B, which is a modification of DELIGHT-2, takes into consideration this heterogeneous effect. The auxiliary code REDEL, a reduction of DELIGHT-B, used in combination with 3 dimensional diffusion code CITATION, is for core burnup calculation with the macro-scopic cross section model. (auth.)

Development of an extended-burnup Mark B design. Second semiannual progress report, January-June 1979

International Nuclear Information System (INIS)

1979-11-01

The immediate goal of the DOE/AP and L/B and W project is to extend the burnup of light water reactor fuel assemblies beyond present limits to 50,000 MWd/mtU batch average burnup. Fuel management plans and fuel designs are being directed to attain the increased burnup limits. Lead-test assemblies of extended-burnup designs will be manufactured, irradiated in a commercial pressurized water reactor, and examined to support extended-burnup fuel cycles. This report, covering the period from January through June 1979, is the second semiannual progress report for the program. Efforts have included analyses of extended-burnup fuel cycles, developed of both annular fuel pellet and segmented rod designs, and design of a nondestructive post-irradiation examination system

Reduction of Defects on Microstructure Aluminium Nitride Using High Temperature Annealing Heat Treatment

Science.gov (United States)

Tanasta, Z.; Muhamad, P.; Kuwano, N.; Norfazrina, H. M. Y.; Unuh, M. H.

2018-03-01

Aluminium Nitride (AlN) is a ceramic 111-nitride material that is used widely as components in functional devices. Besides good thermal conductivity, it also has a high band gap in emitting light which is 6 eV. AlN thin film is grown on the sapphire substrate (0001). However, lattice mismatch between both materials has caused defects to exist along the microstructure of AlN thin films. The defects have affected the properties of Aluminium Nitride. Annealing heat treatment has been proved by the previous researcher to be the best method to improve the microstructure of Aluminium Nitride thin films. Hence, this method is applied at four different temperatures for two hour. The changes of Aluminium Nitride microstructures before and after annealing is observed using Transmission Electron Microscope. It is observed that inversion domains start to occur at temperature of 1500 °C. Convergent Beam Electron Diffraction pattern simulation has confirmed the defects as inversion domain. Therefore, this paper is about to extract the matters occurred during the process of producing high quality Aluminium Nitride thin films and the ways to overcome this problem.

Fission gas release from fuel at high burnup

International Nuclear Information System (INIS)

Meyer, R.O.; Beyer, C.E.; Voglewede, J.C.

1978-03-01

The release of fission gas from fuel pellets at high burnup is reviewed in the context of the safety analysis performed for reactor license applications. Licensing actions are described that were taken to correct deficient gas release models used in these safety analyses. A correction function, which was developed by the Nuclear Regulatory Commission staff and its consultants, is presented. Related information, which includes some previously unpublished data, is also summarized. The report thus provides guidance for the analysis of high burnup gas release in licensing situations

A regime showing anomalous triton burnup in JET

International Nuclear Information System (INIS)

Conroy, S.; Jarvis, O.N.; Sadler, G.; Pillon, M.

1990-01-01

Measurements of triton burnup made at JET in 1989 are in good agreement with a simple classical model of the triton slowing down, for the majority of discharges. For discharges with a long slowing down time (greater than 2 seconds), a much reduced burnup has been observed, suggesting that the tritons undergo diffusion with a diffusion constant of 0.10 m 2 s -1 . Also, the experimental 14 MeV neutron yield is 30% lower than expected for Beryllium limiter discharges. (author) 4 refs., 3 figs

Hot pressing of uranium nitride and mixed uranium plutonium nitride

International Nuclear Information System (INIS)

Chang, J.Y.

1975-01-01

The hot pressing characteristics of uranium nitride and mixed uranium plutonium nitride were studied. The utilization of computer programs together with the experimental technique developed in the present study may serve as a useful purpose of prediction and fabrication of advanced reactor fuel and other high temperature ceramic materials for the future. The densification of nitrides follow closely with a plastic flow theory expressed as: d rho/ dt = A/T(t) (1-rho) [1/1-(1-rho)/sup 2/3/ + B1n (1-rho)] The coefficients, A and B, were obtained from experiment and computer curve fitting. (8 figures) (U.S.)

Burnup calculations for KIPT accelerator driven subcritical facility using Monte Carlo computer codes-MCB and MCNPX

International Nuclear Information System (INIS)

Gohar, Y.; Zhong, Z.; Talamo, A.

2009-01-01

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron accelerator. The neutron source of the subcritical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and electron energy in the range of 100 to 200 MeV. The main functions of the subcritical assembly are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron physics experiments and material structure analyses are planned using this facility. With the 100 KW electron beam power, the total thermal power of the facility is ∼375 kW including the fission power of ∼260 kW. The burnup of the fissile materials and the buildup of fission products reduce continuously the reactivity during the operation, which reduces the neutron flux level and consequently the facility performance. To preserve the neutron flux level during the operation, fuel assemblies should be added after long operating periods to compensate for the lost reactivity. This process requires accurate prediction of the fuel burnup, the decay behavior of the fission produces, and the introduced reactivity from adding fresh fuel assemblies. The recent developments of the Monte Carlo computer codes, the high speed capability of the computer processors, and the parallel computation techniques made it possible to perform three-dimensional detailed burnup simulations. A full detailed three-dimensional geometrical model is used for the burnup simulations with continuous energy nuclear data libraries for the transport calculations and 63-multigroup or one group cross sections libraries for the depletion calculations. Monte Carlo Computer code MCNPX and MCB are utilized for this study. MCNPX transports the electrons and the

Analytical and numerical study of radiation effect up to high burnup in power reactor fuels

International Nuclear Information System (INIS)

Lemes, M; Denis, A; Soba, A

2012-01-01

In the present work the behavior of fuel pellets for power reactors in the high burnup range (average burnup higher than 50 MWd/kgHM) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup, as long as a new microstructure develops, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behaviour. The evolution of porosity in the high burnup structure (HBS) is assumed to be determinant of the retention capacity of the fission gases released by the matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Starting from several works published in the open literature, a model was developed to describe the behaviour and evolution of porosity at local burnup values ranging from 60 to 300 MWd/KgHM. The model is mathematically expressed by a system of non-linear differential equations that take into account the open and closed porosity, the interactions between pores and the free surface and phenomena like pore's coalescence and migration and gas venting. Interactions of different orders between open and closed pores, growth of pores radius by vacancies trapping, the evolution of the pores number density, the internal pressure and over pressure within the pores, the fission gas retained in the matrix and released to the free volume are analyzed. The results of the simulations performed in the present work are in excellent agreement with experimental data available in the open literature and with results calculated by other authors (author)

A guide introducing burnup credit, preliminary version. Contract research

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

It is examined to take burnup credit into account for criticality safety control of facility treating spent fuel. This work is a collection of current technical status of predicting isotopic composition and criticality of spent fuel, points to be specially considered for safety evaluation, and current status of legal affairs for the purpose of applying burnup credit to the criticality safety evaluation of the facility treating spent fuel in Japan. (author)

iBEST: a program for burnup history estimation of spent fuels based on ORIGEN-S

International Nuclear Information System (INIS)

Kim, Do Yeon; Hong, Ser Gi; Ahn, Gil Hoon

2015-01-01

In this paper, we describe a computer program, iBEST (inverse Burnup ESTimator), that we developed to accurately estimate the burnup histories of spent nuclear fuels based on sample measurement data. The burnup history parameters include initial uranium enrichment, burnup, cooling time after discharge from reactor, and reactor type. The program uses algebraic equations derived using the simplified burnup chains of major actinides for initial estimations of burnup and uranium enrichment, and it uses the ORIGEN-S code to correct its initial estimations for improved accuracy. In addition, we newly developed a stable bisection method coupled with ORIGEN-S to correct burnup and enrichment values and implemented it in iBEST in order to fully take advantage of the new capabilities of ORIGEN-S for improving accuracy. The iBEST program was tested using several problems for verification and well-known realistic problems with measurement data from spent fuel samples from the Mihama-3 reactor for validation. The test results show that iBEST accurately estimates the burnup history parameters for the test problems and gives an acceptable level of accuracy for the realistic Mihama-3 problems

Microstructural change and its influence on fission gas release in high burnup UO 2 fuel

Science.gov (United States)

Une, K.; Nogita, K.; Kashibe, S.; Imamura, M.

1992-06-01

The microstructural change of UO 2 fuel pellets (burnup: 6-83 GWd/t), base irradiated under LWR conditions, has been studied by detailed postirradiation examinations. The lattice parameter near the fuel rim in the irradiated UO 2 increased with burnup and appeared to become constant beyond about 50 GWd/t. This lattice dilation was mainly due to the accumulation of radiation induced point defects. Moreover, the dislocation density in the UO 2 matrix developed progressively with burnup, and eventually the tangled dislocations organized many sub-grain boundaries in the highest burnup fuel of 83 GWd/t. This sub-grain structure induced by accumulated radiation damage was compatible in appearance with SEM fractography results which revealed sub-divided grains of sub-micron size in as-fabricated grains. The influence of burnup on 85Kr release from the UO 2 fuels has been examined by means of a postirradiation annealing technique. The higher fractional release of high burnup fuels was mainly due to the burnup dependence of the fractional burst release evolved on temperature ramp. The fractional burst release was represented in terms of the square root of burnup from 6 to 83 GWd/t.

Development and Applications of a Prototypic SCALE Control Module for Automated Burnup Credit Analysis

International Nuclear Information System (INIS)

Gauld, I.C.

2001-01-01

Consideration of the depletion phenomena and isotopic uncertainties in burnup-credit criticality analysis places an increasing reliance on computational tools and significantly increases the overall complexity of the calculations. An automated analysis and data management capability is essential for practical implementation of large-scale burnup credit analyses that can be performed in a reasonable amount of time. STARBUCS is a new prototypic analysis sequence being developed for the SCALE code system to perform automated criticality calculations of spent fuel systems employing burnup credit. STARBUCS is designed to help analyze the dominant burnup credit phenomena including spatial burnup gradients and isotopic uncertainties. A search capability also allows STARBUCS to iterate to determine the spent fuel parameters (e.g., enrichment and burnup combinations) that result in a desired k eff for a storage configuration. Although STARBUCS was developed to address the analysis needs for spent fuel transport and storage systems, it provides sufficient flexibility to allow virtually any configuration of spent fuel to be analyzed, such as storage pools and reprocessing operations. STARBUCS has been used extensively at Oak Ridge National Laboratory (ORNL) to study burnup credit phenomena in support of the NRC Research program

Reactivity effect of spent fuel depending on burn-up history

International Nuclear Information System (INIS)

Hayashi, Takafumi; Suyama, Kenya; Nomura, Yasushi

2001-06-01

It is well known that a composition of spent fuel depends on various parameter changes throughout a burn-up period. In this study we aimed at the boron concentration and its change, the coolant temperature and its spatial distribution, the specific power, the operation mode, and the duration of inspection, because the effects due to these parameters have not been analyzed in detail. The composition changes of spent fuel were calculated by using the burn-up code SWAT, when the parameters mentioned above varied in the range of actual variations. Moreover, to estimate the reactivity effect caused by the composition changes, the criticality calculations for an infinite array of spent fuel were carried out with computer codes SRAC95 or MVP. In this report the reactivity effects were arranged from the viewpoint of what parameters gave more positive reactivity effect. The results obtained through this study are useful to choose the burn-up calculation model when we take account of the burn-up credit in the spent fuel management. (author)

Implementation of burnup credit in spent fuel management systems. Proceedings of an advisory group meeting

International Nuclear Information System (INIS)

1998-04-01

The criticality safety analysis of spent fuel systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system's reactivity. Improved calculational methods allows one to take credit for the reactivity reduction associated with fuel burnup, hence reducing the analysis conservatism while maintaining an adequate criticality safety margin. Motivation for using burnup credit in criticality safety applications is generally based on economic considerations. Although economics may be a primary factor in deciding to use burnup credit, other benefits may be realized. Many of the additional benefits of burnup credit that are not strictly economic, may be considered to contribute to public health and safety, and resource conservation and environmental quality. Interest in the implementation of burnup credit has been shown by many countries. A summary of the information gathered by the IAEA about ongoing activities and regulatory status of burnup credit in different countries is included. Burnup credit implementation introduces new parameters and effects that should be addressed in the criticality analysis (e.g., axial and radial burnup shapes, fuel irradiation history, and others). Analysis of these parameters introduces new variations as well as the uncertainties, that should be considered in the safety assessment of the system. Also, the need arises to validate the isotopic composition that results from a depletion calculation, as well as to extend the current validation range of criticality codes to cover spent fuel. The use of burnup credit implies a verification of the fuel burnup before loading for transport, storage, disposal, or reprocessing each assembly, to make sure that the burnup level achieved complies with the criteria established. Methods and procedures used in different countries are described in this report

Implementation of burnup credit in spent fuel management systems. Proceedings of an advisory group meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-04-01

The criticality safety analysis of spent fuel systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system`s reactivity. Improved calculational methods allows one to take credit for the reactivity reduction associated with fuel burnup, hence reducing the analysis conservatism while maintaining an adequate criticality safety margin. Motivation for using burnup credit in criticality safety applications is generally based on economic considerations. Although economics may be a primary factor in deciding to use burnup credit, other benefits may be realized. Many of the additional benefits of burnup credit that are not strictly economic, may be considered to contribute to public health and safety, and resource conservation and environmental quality. Interest in the implementation of burnup credit has been shown by many countries. A summary of the information gathered by the IAEA about ongoing activities and regulatory status of burnup credit in different countries is included. Burnup credit implementation introduces new parameters and effects that should be addressed in the criticality analysis (e.g., axial and radial burnup shapes, fuel irradiation history, and others). Analysis of these parameters introduces new variations as well as the uncertainties, that should be considered in the safety assessment of the system. Also, the need arises to validate the isotopic composition that results from a depletion calculation, as well as to extend the current validation range of criticality codes to cover spent fuel. The use of burnup credit implies a verification of the fuel burnup before loading for transport, storage, disposal, or reprocessing each assembly, to make sure that the burnup level achieved complies with the criteria established. Methods and procedures used in different countries are described in this report. Refs, figs, tabs.

Ultrasonic measurement of high burn-up fuel elastic properties

International Nuclear Information System (INIS)

Laux, D.; Despaux, G.; Augereau, F.; Attal, J.; Gatt, J.; Basini, V.

2006-01-01

The ultrasonic method developed for the evaluation of high burn-up fuel elastic properties is presented hereafter. The objective of the method is to provide data for fuel thermo-mechanical calculation codes in order to improve industrial nuclear fuel and materials or to design new reactor components. The need for data is especially crucial for high burn-up fuel modelling for which the fuel mechanical properties are essential and for which a wide range of experiments in MTR reactors and high burn-up commercial reactor fuel examinations have been included in programmes worldwide. To contribute to the acquisition of this knowledge the LAIN activity is developing in two directions. First one is development of an ultrasonic focused technique adapted to active materials study. This technique was used few years ago in the EdF laboratory in Chinon to assess the ageing of materials under irradiation. It is now used in a hot cell at ITU Karlsruhe to determine the elastic moduli of high burnup fuels from 0 to 110 GWd/tU. Some of this work is presented here. The second on going programme is related to the qualification of acoustic sensors in nuclear environments, which is of a great interest for all the methods, which work, in a hostile nuclear environment

Preliminary neutronic design of high burnup OTTO cycle pebble bed reactor

International Nuclear Information System (INIS)

Setiadipura, T.; Zuhair; Irwanto, D.

2015-01-01

The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM) loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble. (author)

Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

Directory of Open Access Journals (Sweden)

T. Setiadipura

2015-04-01

Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

TRIGA fuel element burnup determination by measurement and calculation

International Nuclear Information System (INIS)

Zagar, T.; Ravnik, M.; Persic, A.; Jeraj, R.

2000-01-01

To estimate the accuracy of the fuel element burnup calculation different factors influencing the calculation were studied. To cover different aspects of burnup calculations, two in-house developed computer codes were used in calculations. The first (TRIGAP) is based on a one-dimensional two-group diffusion approximation, and the second (TRIGLAV) is based on a two-dimensional four-group diffusion equation. Both codes use WIMSD program with different libraries forunit-cell cross section data calculation. The burnup accumulated during the operating history of the TRIGA reactor at Josef Stefan Institute was calculated for all fuel elements. Elements used in the core during this period were standard SS 8.5% fuel elements, standard SS 12% fuel elements and highly enriched FLIP fuel elements. During the considerable period of operational history, FLIP and standard fuel elements were used simultaneously in mixed cores. (authors)

Metal Nitrides for Plasmonic Applications

DEFF Research Database (Denmark)

Naik, Gururaj V.; Schroeder, Jeremy; Guler, Urcan

2012-01-01

Metal nitrides as alternatives to metals such as gold could offer many advantages when used as plasmonic material. We show that transition metal nitrides can replace metals providing equally good optical performance for many plasmonic applications.......Metal nitrides as alternatives to metals such as gold could offer many advantages when used as plasmonic material. We show that transition metal nitrides can replace metals providing equally good optical performance for many plasmonic applications....

Detailed description and user`s manual of high burnup fuel analysis code EXBURN-I

Energy Technology Data Exchange (ETDEWEB)

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

1997-11-01

EXBURN-I has been developed for the analysis of LWR high burnup fuel behavior in normal operation and power transient conditions. In the high burnup region, phenomena occur which are different in quality from those expected for the extension of behaviors in the mid-burnup region. To analyze these phenomena, EXBURN-I has been formed by the incorporation of such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding oxide layer growth to the basic structure of low- and mid-burnup fuel analysis code FEMAXI-IV. The present report describes in detail the whole structure of the code, models, and materials properties. Also, it includes a detailed input manual and sample output, etc. (author). 55 refs.

Stability characteristics and structural properties of single- and double-walled boron-nitride nanotubes under physical adsorption of Flavin mononucleotide (FMN) in aqueous environment using molecular dynamics simulations

International Nuclear Information System (INIS)

Ansari, R.; Ajori, S.; Ameri, A.

2016-01-01

Graphical abstract: Structural properties and stability characteristics of single- and double-walled boron-nitride nanotubes functionalized with Flavin mononucleotide (FMN) in aqueous environment are investigated employing molecular dynamics simulations. - Highlights: • Structural and buckling analysis of boron-nitride nanotubes under physical adsorption of Flavin mononucleotide (FMN). • Gyration radius increases linearly as the weight percentage of FMN increases. • Presence of water molecules results in more expansion of FMN around BNNTs. • Critical buckling force of functionalized BNNTs is higher than that of pure BNNTs. • The critical strain of functionalized BNNTs is found to be lower than that of pure ones. - Abstract: The non-cytotoxic properties of Boron-nitride nanotubes (BNNTs) and the ability of stable interaction with biomolecules make them so promising for biological applications. In this research, molecular dynamics (MD) simulations are performed to investigate the structural properties and stability characteristics of single- and double-walled BNNTs under physical adsorption of Flavin mononucleotide (FMN) in vacuum and aqueous environments. According to the simulation results, gyration radius increases by rising the weight percentage of FMN. Also, the results demonstrate that critical buckling force of functionalized BNNTs increases in vacuum. Moreover, it is observed that by increasing the weight percentage of FMN, critical force of functionalized BNNTs rises. By contrast, critical strain reduces by functionalization of BNNTs in vacuum. Considering the aqueous environment, it is observed that gyration radius and critical buckling force of functionalized BNNTs increase more considerably than those of functionalized BNNTs in vacuum, whereas the critical strains approximately remain unchanged.

High burnup models in computer code fair

Energy Technology Data Exchange (ETDEWEB)

Dutta, B K; Swami Prasad, P; Kushwaha, H S; Mahajan, S C; Kakodar, A [Bhabha Atomic Research Centre, Bombay (India)

1997-08-01

An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ``Light water reactor fuel rod modelling code evaluation`` and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs.

High burnup models in computer code fair

International Nuclear Information System (INIS)

Dutta, B.K.; Swami Prasad, P.; Kushwaha, H.S.; Mahajan, S.C.; Kakodar, A.

1997-01-01

An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs

KENOREST - A new coupled code system based on KENO and OREST for criticality and burnup inventory calculations

International Nuclear Information System (INIS)

Hesse, U.; Gmal, B.; Voggenberger, Th.; Baleanu, M.; Langenbuch, S.

2001-01-01

The program system KENOREST version 1998 will be presented, which is a useful tool for burnup and reactivity calculations for LWR fuel. The three-dimensional Monte Carlo code KENO-V.a is coupled with the one-dimensional GRS burnup program system OREST-98. The objective is to achieve a better modelling of plutonium and actinide build-up or burnout for advanced heterogeneous fuel assembly designs. Further objectives are directed to reliable calculations of the pin power distributions and of reactor safety parameters including axial and radial rod temperatures for fuel assemblies of modern design. The stand-alone-code KENO-V.a version is used without any changes in the program source. The OREST-98 system was developed to handle multirod problems and additional burnup dependent moderator conditions which can be applied to stretch-out simulations in the reactor. A new interface module RESPEFF between KENO and OREST transforms the 2-d or 3-d KENO flux results to the one-dimensional lattice code OREST in a fully automated manner to maintain reaction rate balance between the codes. First results for assembly multiplication factors, isotope inventories are compared with OECD results. (author)

A burn-up module coupling to an AMPX system

International Nuclear Information System (INIS)

Salvatore Duque, M.; Gomez, S.E.; Patino, N.E.; Abbate, M.J.; Sbaffoni, M.M.

1990-01-01

The Reactors and Neutrons Division of the Bariloche Atomic Center uses the AMPX system for the study of high conversion reactors (HCR). Such system allows to make neutronic calculations from the nuclear data library (ENDF/B-IV). The Nuclear Engineering career of the Balseiro Institute developed and implemented a burn-up module at a μ-cell level (BUM: Burn-up Module) which agrees with the requirement to be coupled to the AMPX system. (Author) [es

Criticality reference benchmark calculations for burnup credit using spent fuel isotopics

International Nuclear Information System (INIS)

Bowman, S.M.

1991-04-01

To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as ''burnup credit,'' the fuel loading capacity of these casks can be increased. One of the difficulties in implementing burnup credit in criticality analyses is that there have been no critical experiments performed with spent fuel which can be used for computer code validation. In lieu of that, a reference problem set of fresh fuel critical experiments which model various conditions typical of light water reactor (LWR) transportation and storage casks has been identified and used in the validation of SCALE-4. This report documents the use of this same problem set to perform spent fuel criticality benchmark calculations by replacing the actual fresh fuel isotopics from the experiments with six different sets of calculated spent fuel isotopics. The SCALE-4 modules SAS2H and CSAS4 were used to perform the analyses. These calculations do not model actual critical experiments. The calculated k-effectives are not supposed to equal unity and will vary depending on the initial enrichment and burnup of the calculated spent fuel isotopics. 12 refs., 11 tabs

Molecular dynamics simulation of nano-indentation of (111) cubic boron nitride with optimized Tersoff potential

Energy Technology Data Exchange (ETDEWEB)

Zhao, Yinbo [College of Aerospace Engineering, Chongqing University, Chongqing 400044 (China); Peng, Xianghe, E-mail: xhpeng@cqu.edu.cn [College of Aerospace Engineering, Chongqing University, Chongqing 400044 (China); State Key Laboratory of Coal Mine Disaster Dynamics and Control, Chongqing University, Chongqing 400044 (China); Fu, Tao; Huang, Cheng; Feng, Chao [College of Aerospace Engineering, Chongqing University, Chongqing 400044 (China); Yin, Deqiang [School of Manufacturing Science and Engineering, Sichuan University, Chengdu 610065 (China); Wang, Zhongchang, E-mail: zcwang@wpi-aimr.tohoku.ac.jp [College of Aerospace Engineering, Chongqing University, Chongqing 400044 (China); Advanced Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai 980-8577 (Japan)

2016-09-30

Highlights: • We optimize Tersoff potential to simulate the cBN better under nanoidentation. • Dislocations slip more easily along <110> and <112> directions on the {111} plane. • Shuffle-set dislocation slip along <112> direction on {111} plane first. • A tetrahedron structure is found in the initial stage of the indentation. - Abstract: We conduct molecular dynamics simulation of nanoindentation on (111) surface of cubic boron nitride and find that shuffle-set dislocations slip along <112> direction on {111} plane at the initial stage of the indentation. The shuffle-set dislocations are then found to meet together, forming surfaces of a tetrahedron. We also find that the surfaces are stacking-fault zones, which intersect with each other, forming edges of stair-rod dislocations along <110> direction. Moreover, we also calculate the generalized stacking fault (GSF) energies along various gliding directions on several planes and find that the GSF energies of the <112>{111} and <110>{111} systems are relatively smaller, indicating that dislocations slip more easily along <110> and <112> directions on the {111} plane.

Characteristics of Au/PZT/TiO2/Nitride/Si structure capacitors with ICP nitride treatments

International Nuclear Information System (INIS)

Min, Hyung Seob; Kim, Tae Ho; Jeon, Chang Bae; Lee, Jae Gab; Kim, Ji Young

2002-01-01

In this study, the characteristics of PZT/TiO 2 ferroelectric gate stack capacitors with Inductively Coupled Plasma (ICP) nitridation were investigated for field effect transistor (FET)-type Ferroelectric Random Access Memory (FeRAM) applications. If a high accumulation capacitance is to be had, the ICP nitridation time needs to be optimized. While a short ICP treatment time results in thermal oxide growth due to lack of nitrogen, a long nitridation time causes a nitride layer which is too thick. Au/PZT(200 nm)/TiO 2 (40 nm)/Nitride/Si (MeFINS) structure capacitors show a memory window (ΔV) of 1.6 V under ±3-V operation while Au/PZT(200 nm)/TiO 2 (40 nm)/Si (MeFIS) capacitors without nitride treatment exhibit a small memory window of 0.6 V. At the same time, the capacitance of the MeFINS device is almost twice that of the MeFIS capacitor. This result implies that the ICP nitride treatment suppresses the formation of a low dielectric constant interfacial SiO x layer and alleviates the series capacitance problem

Communication: Water on hexagonal boron nitride from diffusion Monte Carlo

Energy Technology Data Exchange (ETDEWEB)

Al-Hamdani, Yasmine S.; Ma, Ming; Michaelides, Angelos, E-mail: angelos.michaelides@ucl.ac.uk [Thomas Young Centre and London Centre for Nanotechnology, 17–19 Gordon Street, London WC1H 0AH (United Kingdom); Department of Chemistry, University College London, 20 Gordon Street, London WC1H 0AJ (United Kingdom); Alfè, Dario [Thomas Young Centre and London Centre for Nanotechnology, 17–19 Gordon Street, London WC1H 0AH (United Kingdom); Department of Earth Sciences, University College London, Gower Street, London WC1E 6BT (United Kingdom); Lilienfeld, O. Anatole von [Institute of Physical Chemistry and National Center for Computational Design and Discovery of Novel Materials, Department of Chemistry, University of Basel, Klingelbergstrasse 80, CH-4056 Basel (Switzerland); Argonne Leadership Computing Facility, Argonne National Laboratories, 9700 S. Cass Avenue Argonne, Lemont, Illinois 60439 (United States)

2015-05-14

Despite a recent flurry of experimental and simulation studies, an accurate estimate of the interaction strength of water molecules with hexagonal boron nitride is lacking. Here, we report quantum Monte Carlo results for the adsorption of a water monomer on a periodic hexagonal boron nitride sheet, which yield a water monomer interaction energy of −84 ± 5 meV. We use the results to evaluate the performance of several widely used density functional theory (DFT) exchange correlation functionals and find that they all deviate substantially. Differences in interaction energies between different adsorption sites are however better reproduced by DFT.

Experimental and computational study of nitride precipitation in a CrMnN austenitic stainless steel

International Nuclear Information System (INIS)

Pettersson, Niklas; Frisk, Karin; Fluch, Rainer

2017-01-01

The austenitic CrMnN stainless steels are high-strength, tough, and non-magnetic, and are used in oil field applications. The steels have high alloying contents, and precipitation of Cr-nitrides and/or intermetallic phases can occur when cooling through the temperature region 950–700 °C. The nitride precipitates appear in the grain boundaries but can be difficult to observe in the microstructure due to their small size. However, there is an effect of precipitation on corrosion and impact strength and a modelling approach to predict precipitation is valuable for alloy and process development. In the present work precipitation simulations were applied to a CrMnN steel composition, and coupled to experimental investigations after heat treatments at 700 and 800 °C. The early stages, with short heat-treatment times, were studied. The simulations were performed using TC-PRISMA, a software for calculation of multiphase precipitation kinetics, using multicomponent nucleation and growth models. Dedicated thermodynamic and kinetic databases were used for the simulations. The main precipitate was identified by experiments and simulations to be the Cr 2 N nitride, and the precipitation during isothermal heat treatments was investigated. Isothermal precipitation diagrams are simulated, and the influence of precipitation kinetics on toughness is discussed.

Experimental and computational study of nitride precipitation in a CrMnN austenitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Pettersson, Niklas, E-mail: niklas.pettersson@swerea.se [Swerea KIMAB AB, P.O. Box 7047, 164 07 Kista (Sweden); Frisk, Karin [Swerea KIMAB AB, P.O. Box 7047, 164 07 Kista (Sweden); Fluch, Rainer [Böhler Edelstahl Gmbh, Mariazeller Strasse 25, 8605 Kapfenberg (Austria)

2017-01-27

The austenitic CrMnN stainless steels are high-strength, tough, and non-magnetic, and are used in oil field applications. The steels have high alloying contents, and precipitation of Cr-nitrides and/or intermetallic phases can occur when cooling through the temperature region 950–700 °C. The nitride precipitates appear in the grain boundaries but can be difficult to observe in the microstructure due to their small size. However, there is an effect of precipitation on corrosion and impact strength and a modelling approach to predict precipitation is valuable for alloy and process development. In the present work precipitation simulations were applied to a CrMnN steel composition, and coupled to experimental investigations after heat treatments at 700 and 800 °C. The early stages, with short heat-treatment times, were studied. The simulations were performed using TC-PRISMA, a software for calculation of multiphase precipitation kinetics, using multicomponent nucleation and growth models. Dedicated thermodynamic and kinetic databases were used for the simulations. The main precipitate was identified by experiments and simulations to be the Cr{sub 2}N nitride, and the precipitation during isothermal heat treatments was investigated. Isothermal precipitation diagrams are simulated, and the influence of precipitation kinetics on toughness is discussed.

Properties of minor actinide nitrides

International Nuclear Information System (INIS)

Takano, Masahide; Itoh, Akinori; Akabori, Mitsuo; Arai, Yasuo; Minato, Kazuo

2004-01-01

The present status of the research on properties of minor actinide nitrides for the development of an advanced nuclear fuel cycle based on nitride fuel and pyrochemical reprocessing is described. Some thermal stabilities of Am-based nitrides such as AmN and (Am, Zr)N were mainly investigated. Stabilization effect of ZrN was cleary confirmed for the vaporization and hydrolytic behaviors. New experimental equipments for measuring thermal properties of minor actinide nitrides were also introduced. (author)

Laser Nitriding of the Newly Developed Ti-20Nb-13Zr at.% Biomaterial Alloy to Enhance Its Mechanical and Corrosion Properties in Simulated Body Fluid

Science.gov (United States)

Hussein, M. A.; Kumar, A. Madhan; Yilbas, Bekir S.; Al-Aqeeli, N.

2017-11-01

Despite the widespread application of Ti alloy in the biomedical field, surface treatments are typically applied to improve its resistance to corrosion and wear. A newly developed biomedical Ti-20Nb-13Zr at.% alloy (TNZ) was laser-treated in nitrogen environment to improve its surface characteristics with corrosion protection performance. Surface modification of the alloy by laser was performed through a Nd:YAG laser. The structural and surface morphological alterations in the laser nitrided layer were investigated by XRD and a FE-SEM. The mechanical properties have been evaluated using nanoindentation for laser nitride and as-received samples. The corrosion protection behavior was estimated using electrochemical corrosion analysis in a physiological medium (SBF). The obtained results revealed the production of a dense and compact film of TiN fine grains (micro-/nanosize) with 9.1 µm below the surface. The mechanical assessment results indicated an improvement in the modulus of elasticity, hardness, and resistance of the formed TiN layer to plastic deformation. The electrochemical analysis exhibited that the surface protection performance of the laser nitrided TNZ substrates in the SBF could be considerably enhanced compared to that of the as-received alloy due to the presence of fine grains in the TiN layer resulting from laser nitriding. Furthermore, the untreated and treated Ti-20Nb-13Zr alloy exhibited higher corrosion resistance than the CpTi and Ti6Al4V commercial alloys. The improvements in the surface hardness and corrosion properties of Ti alloy in a simulated body obtained using laser nitriding make this approach a suitable candidate for enhancing the properties of biomaterials.

Taking burnup credit for interim storage and transportation system for BWR fuels

International Nuclear Information System (INIS)

Yoshioka, Ken-ichi; Ando, Y.; Kumanomido, H.; Sasaki, T.; Mitsuhashi, I.; Ueda, M.

2001-01-01

In order to establish a realistic burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of several modules such as TGBLA, ORIGEN, CITATION, MCNP, and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. A compact measurement system for a fuel assembly has been being developed to meet requirements for the burnup determination, the neutron emission-rate evaluation, and the nuclear materials management. For a spent MOX fuel, a neutron emission rate measurement method has been being developed. The system consists of Cd-Te detectors and / or fission chambers. Some model calculations were carried out for the latest design BWR fuels. The effect of taking burnup credit for a transportation cask is shown. (authors)

Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

International Nuclear Information System (INIS)

Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi

2014-01-01

Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided

THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

Directory of Open Access Journals (Sweden)

MEHMET E. KORKMAZ

2014-06-01

Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

2D to 3D transition of polymeric carbon nitride nanosheets

Energy Technology Data Exchange (ETDEWEB)

Chamorro-Posada, Pedro [Dpto. de Teoría de la Señal y Comunicaciones e IT, Universidad de Valladolid, ETSI Telecomunicación, Paseo Belén 15, 47011 Valladolid (Spain); Vázquez-Cabo, José [Dpto. de Teoría de la Señal y Comunicaciones, Universidad de Vigo, ETSI Telecomunicación, Lagoas Marcosende s/n, Vigo (Spain); Sánchez-Arévalo, Francisco M. [Instituto de Investigaciones en Materiales (IIM), Universidad Nacional Autónoma de México, Apdo. Postal 70–360, Cd. Universitaria, México D.F. 04510 (Mexico); Martín-Ramos, Pablo [Dpto. de Teoría de la Señal y Comunicaciones e IT, Universidad de Valladolid, ETSI Telecomunicación, Paseo Belén 15, 47011 Valladolid (Spain); Laboratorio de Materiales Avanzados (Advanced Materials Laboratory) ETSIIAA, Universidad de Valladolid, Avenida de Madrid 44, 34004 Palencia (Spain); Martín-Gil, Jesús; Navas-Gracia, Luis M. [Laboratorio de Materiales Avanzados (Advanced Materials Laboratory) ETSIIAA, Universidad de Valladolid, Avenida de Madrid 44, 34004 Palencia (Spain); Dante, Roberto C., E-mail: rcdante@yahoo.com [Laboratorio de Materiales Avanzados (Advanced Materials Laboratory) ETSIIAA, Universidad de Valladolid, Avenida de Madrid 44, 34004 Palencia (Spain)

2014-11-15

The transition from a prevalent turbostratic arrangement with low planar interactions (2D) to an array of polymeric carbon nitride nanosheets with stronger interplanar interactions (3D), occurring for samples treated above 650 °C, was detected by terahertz-time domain spectroscopy (THz-TDS). The simulated 3D material made of stacks of shifted quasi planar sheets composed of zigzagged polymer ribbons, delivered a XRD simulated pattern in relatively good agreement with the experimental one. The 2D to 3D transition was also supported by the simulation of THz-TDS spectra obtained from quantum chemistry calculations, in which the same broad bands around 2 THz and 1.5 THz were found for 2D and 3D arrays, respectively. This transition was also in accordance with the tightening of the interplanar distance probably due to an interplanar π bond contribution, as evidenced also by a broad absorption around 2.6 eV in the UV–vis spectrum, which appeared in the sample treated at 650 °C, and increased in the sample treated at 700 °C. The band gap was calculated for 1D and 2D cases. The value of 3.374 eV for the 2D case is, within the model accuracy and precision, in a relative good agreement with the value of 3.055 eV obtained from the experimental results. - Graphical abstract: 2D lattice mode vibrations and structural changes correlated with the so called “2D to 3D transition”. - Highlights: • A 2D to 3D transition has been detected for polymeric carbon nitride. • THz-TDS allowed us to discover and detect the 2D to 3D transition of polymeric carbon nitride. • We propose a structure for polymeric carbon nitride confirming it with THz-TDS.

2D to 3D transition of polymeric carbon nitride nanosheets

International Nuclear Information System (INIS)

Chamorro-Posada, Pedro; Vázquez-Cabo, José; Sánchez-Arévalo, Francisco M.; Martín-Ramos, Pablo; Martín-Gil, Jesús; Navas-Gracia, Luis M.; Dante, Roberto C.

2014-01-01

The transition from a prevalent turbostratic arrangement with low planar interactions (2D) to an array of polymeric carbon nitride nanosheets with stronger interplanar interactions (3D), occurring for samples treated above 650 °C, was detected by terahertz-time domain spectroscopy (THz-TDS). The simulated 3D material made of stacks of shifted quasi planar sheets composed of zigzagged polymer ribbons, delivered a XRD simulated pattern in relatively good agreement with the experimental one. The 2D to 3D transition was also supported by the simulation of THz-TDS spectra obtained from quantum chemistry calculations, in which the same broad bands around 2 THz and 1.5 THz were found for 2D and 3D arrays, respectively. This transition was also in accordance with the tightening of the interplanar distance probably due to an interplanar π bond contribution, as evidenced also by a broad absorption around 2.6 eV in the UV–vis spectrum, which appeared in the sample treated at 650 °C, and increased in the sample treated at 700 °C. The band gap was calculated for 1D and 2D cases. The value of 3.374 eV for the 2D case is, within the model accuracy and precision, in a relative good agreement with the value of 3.055 eV obtained from the experimental results. - Graphical abstract: 2D lattice mode vibrations and structural changes correlated with the so called “2D to 3D transition”. - Highlights: • A 2D to 3D transition has been detected for polymeric carbon nitride. • THz-TDS allowed us to discover and detect the 2D to 3D transition of polymeric carbon nitride. • We propose a structure for polymeric carbon nitride confirming it with THz-TDS

Improvements on burnup chain model and group cross section library in the SRAC system

International Nuclear Information System (INIS)

Akie, Hiroshi; Okumura, Keisuke; Takano, Hideki; Ishiguro, Yukio; Kaneko, Kunio.

1992-01-01

Data and functions of the cell burnup calculation of the SRAC system were revised to improve mainly the accuracy of the burnup calculation of high conversion light water reactors (HCLWRs). New burnup chain models were developed in order to treat fission products (FPs) and actinide nuclides in detail. Group cross section library, SRACLIB-JENDL2, was generated based on JENDL-2 nuclear data file. In generating this library, emphasis was placed on FPs and actinides. Also revised were the data such as the average energy release per fission for various actinides. These improved data were verified by performing the burnup analysis of PWR spent fuels. Some new functions were added to the SRAC system for the convenience to yield macroscopic cross sections used in the core burnup process. (author)

Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Okubo, Tsutomu

2011-01-01

Highlights: → Pu isotopic composition of fuel affects FBR core nuclear characteristics very much. → Spent fuel compositions of next generation LWRs with burnup of 70 GWd/t were obtained. → Pu isotopic composition and amount in the spent fuel with 70 GWd/t were evaluated. → Spectral shift rods of high burnup BWR increases the fissile Pu fraction of spent fuel. → Wide fuel rod pitch of high burnup PWR lowers the fissile Pu fraction of spent fuel. - Abstract: The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs. The HB-BWR employs spectral shift rods and the neutron spectrum is shifted through the operation cycle. The weight fraction of fissile plutonium (Puf) isotopes to the total plutonium in HB-BWR spent fuel after 5 years cooling is 62%, which is larger than that of conventional BWRs with average burnup of 45 GWd/t, because of the spectral shift operation. The amount of Pu produced in the HB-BWR is also larger than that produced in a conventional BWR. The HB-PWR uses a wider pitch 17 x 17 fuel rod assembly to optimize neutron slowing down. The Puf fraction of HB-PWR spent fuel after 5 years cooling is 56%, which is smaller than that of conventional PWRs with average burnup of 49 GWd/t, mainly because of the wider pitch. The amount of Pu produced in the HB-PWR is also smaller than that in conventional PWRs.

Analysis on burn-up behaviors for accelerator-driven sub-critical facility

International Nuclear Information System (INIS)

Liu Guisheng; Zhao Zhixiang; Zhang Baocheng; Shen Qinbiao; Ding Dazhao

2000-01-01

An analysis is performed on burn-up behaviors for accelerator-driven sub-critical reactor by means of the code PASC-1 for neutronics calculation, the code CBURN for burn-up calculation and 44 group constants is processed by CENDL-2 and ENDF/B-6 using NJOY-91.91

Dynamic response of multiwall boron nitride nanotubes subjected to ...

Indian Academy of Sciences (India)

Page 1 ... 1. Introduction. Boron nitride nanotubes (BNNTs) are like carbon nanotubes. (CNTs) in structure in which carbon atoms are replaced by alternate boron and nitrogen atoms. Thus, BNNTs demon- ... istic analyser for intermediate landing situation of inserted mass.15 Also, a macroscopic continuum simulation is sug-.

Current studies related to the use of burnup credit in France

International Nuclear Information System (INIS)

Raby, Jerome; Lavarenne, Caroline; Barreau, Anne; Riffard, Cecile; Roque, Benedicte; Bioux, Philippe; Doucet, Michel; Guillou, Eric; Leka, Georges; Toubon, Herve

2003-01-01

In order to avoid criticality risks, a large number of facilities using spent fuels have been designed considering the fuel as fresh. This choice has obviously led to considerable safety margins. In the early 80's, a method was accepted by the French Safety Authorities allowing to consider the changes in the fuel composition during the depletion with some very pessimistic hypothesis: only actinides were considered and the amount of burnup used in the studies was equal to the mean burnup in the 50-least-irradiated centimeters. As many facilities still want to optimize their processes (e.g. transportation, storage, fuel reprocessing), the main companies involved in the French nuclear industry, researchers and IRSN set up a Working Group in order to study the way burnup could be taken into account in the criticality calculations, considering some fission products and a more realistic axial profile of burnup. The first of this article introduces the current French method used to take burnup into account in the criticality studies. The second part is devoted to the studies achieved by the Working Group to improve this method, especially concerning the consideration of the neutron absorption of some fission products and of an axial profile of burnup: for that purpose, some results are presented related to the steps of the process like the depletion calculations, the definition of an axial profile and the criticality calculation. In the third part, some results (keff) obtained with fission products and an axial profile are compared to those obtained with the current one. The conclusions presented are related to the present state of knowledge and may differ from the final conclusions of the Working Group. (author)

Development and benchmark verification of a parallelized Monte Carlo burnup calculation program MCBMPI

International Nuclear Information System (INIS)

Yang Wankui; Liu Yaoguang; Ma Jimin; Yang Xin; Wang Guanbo

2014-01-01

MCBMPI, a parallelized burnup calculation program, was developed. The program is modularized. Neutron transport calculation module employs the parallelized MCNP5 program MCNP5MPI, and burnup calculation module employs ORIGEN2, with the MPI parallel zone decomposition strategy. The program system only consists of MCNP5MPI and an interface subroutine. The interface subroutine achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, data exchanging with MCNP5MPI. Also, the program was verified with the Pressurized Water Reactor (PWR) cell burnup benchmark, the results showed that it's capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)

Fission-gas release in fuel performing to extended burnups in Ontario Hydro nuclear generating stations

International Nuclear Information System (INIS)

Floyd, M.R.; Novak, J.; Truant, P.T.

1992-06-01

The average discharge burnup of CANDU fuel is about 200 MWh/kgU. A significant number of 37-element bundles have achieved burnups in excess of 400 MWh/kgU. Some of these bundles have experienced failures related to their extended operation. To date, hot-cell examinations have been performed on fuel elements from nine 37-element bundles irradiated in Bruce NGS-A that have burnups in the range of 300-800 MWh/kgU. 1 Most of these have declining power histories from peak powers of up to 59 kW/m. Fission-gas releases of up to 26% have been observed and exhibit a strong dependence on fuel power. This obscures any dependence on burnup. The extent of fission-gas release at extended burnups was not predicted by low-burnup code extrapolations. This is attributed primarily to a reduction in fuel thermal conductivity which results in elevated operating temperatures. Reduced conductivity is due, at least in part, to the buildup of fission products in the fuel matrix. Some evidence of hyperstoichiometry exists, although this needs to be further investigated along with any possible relation to CANLUB graphite coating behaviour and sheath oxidation. Residual tensile sheath strains of up to 2% have been observed and can be correlated with fuel power/fission-gas release. SCC 2 -related defects have been observed in the sheath and endcaps of elements from bundles experiencing declining power histories to burnups in excess of 500 MWh/kgU. This indicates that the current recommended burnup limit of 450 MWh/kgU is justified. SCC-related defects have also been observed in ramped bundles having burnups < 450 MWh/kgU. Hence, additional guidelines are in place for power ramping extended-burnup fuel

Fission gas release and pellet microstructure change of high burnup BWR fuel

International Nuclear Information System (INIS)

Itagaki, N.; Ohira, K.; Tsuda, K.; Fischer, G.; Ota, T.

1998-01-01

UO 2 fuel, with and without Gadolinium, irradiated for three, five, and six irradiation cycles up to about 60 GWd/t pellet burnup in a commercial BWR were studied. The fission gas release and the rim effect were investigated by the puncture test and gas analysis method, OM (optical microscope), SEM (scanning electron microscope), and EPMA (electron probe microanalyzer). The fission gas release rate of the fuel rods irradiated up to six cycles was below a few percent; there was no tendency for the fission gas release to increase abruptly with burnup. On the other hand, microstructure changes were revealed by OM and SEM examination at the rim position with burnup increase. Fission gas was found depleted at both the rim position and the pellet center region using EPMA. There was no correlation between the fission gas release measured by the puncture test and the fission gas depletion at the rim position using EPMA. However, the depletion of fission gas in the center region had good correlation with the fission gas release rate determined by the puncture test. In addition, because the burnup is very large at the rim position of high burnup fuel and also due to the fission rate of the produced Pu, the Xe/Kr ratio at the rim position of high burnup fuel is close to the value of the fission yield of Pu. The Xe/Kr ratio determined by the gas analysis after the puncture test was equivalent to the fuel average but not to the pellet rim position. From the results, it was concluded that fission gas at the rim position was released from the UO 2 matrix in high burnup, however, most of this released fission gas was held in the porous structure and not released from the pellet to the free volume. (author)

Experimental and theoretical burnup investigations on model arrangements with solid burnable poisons

International Nuclear Information System (INIS)

Ahlf, J.; Anders, D.; Greim, L.; Knoth, J.; Kolb, M.; Mittelstaedt, B.; Mueller, A.; Schwenke, H.

1975-01-01

It is the scope of the two experiments here to improve the methods for computation and measurement as well as the experimental technique appropriate to predict the burnable poison rod burn-up with sufficient accuracy. In the first experiment two nine-rod bundles in a 3 x 3 arrangement are irradiated during several irradiation periods in the research reactor Geesthacht. Each bundle consists of eight outer rods containing fuel and one inner rod containing poison (B 10 or Cd 113). The burn-up of the fuel and the burnable poison is measured by non-destructive methods after each irradiation period and then compared with results of a burn-up calculation. In the second experiment two poison rods with different cadmium concentrations and one rod containing boron are irradiated during several irradiation periods in the research reactor Geesthacht. The burn-up is determined after each irradiation period by reactivity measurements and its result compared to computed effective absorption cross-sections of the rods by aid of a calibration curve. For both experiments the experimental and theoretical results for the poison burn-up are found to be within the error limits of the measurements. (orig.) [de

Experimental and theoretical investigations on solid burnable poison burnup of model arrangements

International Nuclear Information System (INIS)

Ahlf, J.; Anders, D.; Greim, L.; Knoth, J.; Kolb, M.; Mittelstaedt, B.; Mueller, A.; Schwenke, H.

1975-01-01

It is the scope of the two experiments reported here to improve the methods for computation and measurement as well as the experimental technique appropriate to predict the burnable poison rod burn-up with sufficient accuracy. In the first experiment two nine-rod bundles in a 3 x 3 arrangement are irradiated during several irradiation periods in the research reactor Geesthacht. Each bundle consists of eight outer rods containing fuel and one inner rod containing poison (B 10 or Cd 113). The burn-up of the fuel and the burnable poison is measured by non-destructive methods after each irradiation period and then compared with results of a burn-up calculation. In the second experiment two poison rods with different cadmium concentrations and one rod containing boron are irradiated during several irradiation periods in the research reactor Geesthacht. The burn-up is determined after each irradiation period by reactivity measurements and its result compared to computed effective absorption cross-sections of the rods by aid of a calibration curve. For both experiments the experimental and theoretical results for the poison burn-up are found to be within the error limits of the measurements. (orig.) [de

Extended burnup with SEU fuel in Atucha-1 NPP

International Nuclear Information System (INIS)

Alvarez, L.; Casario, J.; Fink, J.; Perez, R.; Higa, M.

2002-01-01

Atucha-1 is a Pressurized Heavy Water Reactor originally fuelled with natural uranium. Fuel Assemblies consist of 36 fuel rods and the active length is 5300 mm. The total length of the fuel assembly is about 6 m. The average discharge burnup of natural UO 2 fuel is 5900 MWd/tU. After the deregulation of the Argentine electricity market there was an important incentive to reduce the impact of fuel cost on the cost of generation. To keep the competitiveness of the nuclear energy against another sources of electricity it was necessary to reduce the cost of the nuclear fuel. With this objective a program to introduce SEU (0.85 % 235 U) fuel in Atucha-1 was launched in 1993. As a result of this program the average SEU fuel discharge burnup increased to more than 11000 MWd/tU. The first SEU fuels were introduced in Atucha-1 in 1995 and, in the present stage of the program, 71% of core positions are loaded with this type of fuel. This paper describes key aspects of Atucha-1 fuel design and their relevance limiting the burnup extension and shows relevant data regarding the SEU in-reactor performance. At the present time 125 SEU Fuel Assemblies have been irradiated without failures associated with the extended burnup or unfavorable influences on the operation of the power station. (author)

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

Burnup measurements with the Los Alamos fork detector

International Nuclear Information System (INIS)

Bosler, G.E.; Rinard, P.M.

1991-01-01

The fork detector system can determine the burnup of spent-fuel assemblies. It is a transportable instrument that can be mounted permanently in a spent-fuel pond near a loading area for shipping casks, or be attached to the storage pond bridge for measurements on partially raised spent-fuel assemblies. The accuracy of the predicted burnup has been demonstrated to be as good as 2% from measurements on assemblies in the United States and other countries. Instruments have also been developed at other facilities throughout the world using the same or different techniques, but with similar accuracies. 14 refs., 2 figs., 2 tabs

Findings of an international study on burnup credit

International Nuclear Information System (INIS)

Brady, M.C.; Takano, M.; Okuno, H.; DeHart, M.D.; Nouri, A.

1996-01-01

Findings from a four year study by an international benchmarking group in the comparison of computational methods for evaluating burnup credit in criticality safety analyses are presented in this paper. Approximately 20 participants from 11 countries have provided results for most problems. Four detailed benchmark problems for Pressurized Water Reactor (PWR) fuel have been completed and are summarized in this paper. Preliminary results from current work addressing burnup credit for Boiling Water Reactor (BWR) fuel will also be discussed as well as planned activities for additional benchmarks including Mixed-Oxide (MOX) fuels, subcritical benchmarks, international databases, and other activities

Burnup measurements on spent fuel elements of the RP-10 research reactor

Energy Technology Data Exchange (ETDEWEB)

Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

High Burnup Fuel: Implications and Operational Experience. Proceedings of a Technical Meeting

International Nuclear Information System (INIS)

2016-08-01

This publication reports on the outcome of a technical meeting on high burnup fuel experience and economics, held in Buenos Aires, Argentina in 2013. The purpose of the meeting was to revisit and update the current operational experience and economic conditions associated with high burnup fuel. International experts with significant experience in experimental programmes on high burnup fuel discussed and evaluated physical limitations at pellet, cladding and structural component levels, with a wide focus including fabrication, core behaviour, transport and intermediate storage for most types of commercial nuclear power plants

Burnup measurements on spent fuel elements of the RP-10 research reactor

International Nuclear Information System (INIS)

Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro

2011-01-01

This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using 137 Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

Burnup verification measurements on spent fuel assemblies at Arkansas Nuclear One

International Nuclear Information System (INIS)

Ewing, R.I.

1995-01-01

Burnup verification measurements have been performed using the Fork system at Arkansas Nuclear One, Units 1 and 2, operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an internal calibration for the system in the form of a power law determined by a least squares fit to the neutron data. The values of the exponent in the power laws were 3.83 and 4.35 for Units 1 and 2, respectively. The average deviation of the reactor burnup records from the calibration determined from the measurements is a measure of the random error in the burnup records. The observed average deviations were 2.7% and 3.5% for assemblies at Units 1 and 2, respectively, indicating a high degree of consistency in the reactor records. Two non-standard assemblies containing neutron sources were studied at Unit 2. No anomalous measurements were observed among the standard assemblies at either Unit. The effectiveness of the Fork system for verification of reactor records is due to the sensitivity of the neutron yield to burnup, the self-calibration generated by a series of measurements, the redundancy provided by three independent detection systems, and the operational simplicity and flexibility of the design

OECD/NEA burnup credit criticality benchmarks phase IIIB: Burnup calculations of BWR fuel assemblies for storage and transport

International Nuclear Information System (INIS)

Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya

2002-02-01

The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of ±10% relative to the average, although some results, esp. 155 Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k ∞ also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

OECD/NEA burnup credit criticality benchmarks phase IIIB. Burnup calculations of BWR fuel assemblies for storage and transport

Energy Technology Data Exchange (ETDEWEB)

Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-02-01

The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

Development of a set of benchmark problems to verify numerical methods for solving burnup equations

International Nuclear Information System (INIS)

Lago, Daniel; Rahnema, Farzad

2017-01-01

Highlights: • Description transmutation chain benchmark problems. • Problems for validating numerical methods for solving burnup equations. • Analytical solutions for the burnup equations. • Numerical solutions for the burnup equations. - Abstract: A comprehensive set of transmutation chain benchmark problems for numerically validating methods for solving burnup equations was created. These benchmark problems were designed to challenge both traditional and modern numerical methods used to solve the complex set of ordinary differential equations used for tracking the change in nuclide concentrations over time due to nuclear phenomena. Given the development of most burnup solvers is done for the purpose of coupling with an established transport solution method, these problems provide a useful resource in testing and validating the burnup equation solver before coupling for use in a lattice or core depletion code. All the relevant parameters for each benchmark problem are described. Results are also provided in the form of reference solutions generated by the Mathematica tool, as well as additional numerical results from MATLAB.

ABB high burnup fuel

International Nuclear Information System (INIS)

Andersson, S.; Helmersson, S.; Nilsson, S.; Jourdain, P.; Karlsson, L.; Limback, M.; Garde, A.M.

1999-01-01

Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both PWR and BWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter proven to meet the 6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10 x 10 fuel, where ABB is the only vendor to date with batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of PWR and BWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its utility customers. This paper provides an overview of recent fuel performance and reliability experience at ABB. Selected development and validation activities for PWR and BWR fuel are presented, for which the ABB test facilities in Windsor (TF-2 loop, mechanical test laboratory) and Vaesteras (FRIGG, BURE) are essential. (authors)

Observations on the CANDLE burn-up in various geometries

International Nuclear Information System (INIS)

Seifritz, W.

2007-01-01

We have looked at all geometrical conditions under which an auto catalytically propagating burnup wave (CANDLE burn-up) is possible. Thereby, the Sine Gordon equation finds a new place in the burn-up theory of nuclear fission reactors. For a practical reactor design the axially burning 'spaghetti' reactor and the azimuthally burning 'pancake' reactor, respectively, seem to be the most promising geometries for a practical reactor design. Radial and spherical burn-waves in cylindrical and spherical geometry, respectively, are principally impossible. Also, the possible applicability of such fission burn-waves on the OKLO-phenomenon and the GEOREACTOR in the center of Earth, postulated by Herndon, is discussed. A fast CANDLE-reactor can work with only depleted uranium. Therefore, uranium mining and uranium-enrichment are not necessary anymore. Furthermore, it is also possible to dispense with reprocessing because the uranium utilization factor is as high as about 40%. Thus, this completely new reactor type can open a new era of reactor technology

Modelling of some high burnup phenomena in nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Forsberg, K; Lindstroem, F; Massih, A R [ABB Atom AB, Vaesteraas (Sweden)

1997-08-01

In this paper the results of some modelling efforts carried out by ABB Atom to describe certain light water reactor fuel high burnup effects are presented. In particular the degradation of fuel thermal conductivity with burnup and its impact on fuel temperature is briefly discussed. The formation of a porous rim and its effect on a thermal fission gas release has been modelled and the model has been used to predict the release of pressurized water reactor fuel rods that were operated at low power densities. Furthermore, a mathematical model which combines the diffusion and re-solution controlled thermal release with grain boundary movement has been briefly described. The model is used to compare release with diffusion only and release caused by diffusion and grain boundary sweeping (due to grain growth). Finally, analytical expressions are obtained for the calculation of fuel stoichiometry as a function of burnup. (author). 20 refs, 10 figs, 1 tab.

Preparation of higher-actinide burnup and cross section samples

International Nuclear Information System (INIS)

Adair, H.L.; Kobisk, E.H.; Quinby, T.C.; Thomas, D.K.; Dailey, J.M.

1981-01-01

A joint research program involving the United States and the United Kingdom was instigated about four years ago for the purpose of studying burnup of higher actinides using in-core irradiation in the fast reactor at Dounreay, Scotland. Simultaneously, determination of cross sections of a wide variety of higher actinide isotopes was proposed. Coincidental neutron flux and energy spectral measurements were to be made using vanadium encapsulated dosimetry materials in the immediate region of the burnup and cross section samples. The higher actinide samples chosen for the burnup study were 241 Am and 244 Cm in the forms of Am 2 O 3 , Cm 2 O 3 , and Am 6 Cm(RE) 7 O 21 , where (RE) represents a mixture of lanthanide sesquioxides. It is the purpose of this paper to describe technology development and its application in the preparation of the fuel specimens and the cross section specimens that are being used in this cooperative program

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

Linear and nonlinear characterization of low-stress high-confinement silicon-rich nitride waveguides.

Science.gov (United States)

Krückel, Clemens J; Fülöp, Attila; Klintberg, Thomas; Bengtsson, Jörgen; Andrekson, Peter A; Torres-Company, Víctor

2015-10-05

In this paper we introduce a low-stress silicon enriched nitride platform that has potential for nonlinear and highly integrated optics. The manufacturing process of this platform is CMOS compatible and the increased silicon content allows tensile stress reduction and crack free layer growth of 700 nm. Additional benefits of the silicon enriched nitride is a measured nonlinear Kerr coefficient n(2) of 1.4·10(-18) m(2)/W (5 times higher than stoichiometric silicon nitride) and a refractive index of 2.1 at 1550 nm that enables high optical field confinement allowing high intensity nonlinear optics and light guidance even with small bending radii. We analyze the waveguide loss (∼1 dB/cm) in a spectrally resolved fashion and include scattering loss simulations based on waveguide surface roughness measurements. Detailed simulations show the possibility for fine dispersion and nonlinear engineering. In nonlinear experiments we present continuous-wave wavelength conversion and demonstrate that the material does not show nonlinear absorption effects. Finally, we demonstrate microfabrication of resonators with high Q-factors (∼10(5)).

Polynomial expansion methodology for microscopic cross sections to use in spatial burnup calculations

International Nuclear Information System (INIS)

Conti Filho, P.; Oliveira Barroso, A.C. de

1985-01-01

It was developed a computer code to generate polynomial coefficients which represent homogenized microscopic cross sections in function of the local accumulated burnup and concentration of soluble boron, presented in fuel element, for each step of burnup reactor. Afterward, it was developed a coupling between LEOPARD-GERADOR DE POLINOMIOS - CITATION computer codes to interpret and build homogenized microscopic cross sections according with local characteristics of each fuel element during the burnup calculation of reactor core. (M.C.K.) [pt

Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

International Nuclear Information System (INIS)

Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L.; Saito, M.

2003-01-01

Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, 237 Np, 238 Pu, 231 Pa, 232 U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations

Development of a BWR core burn-up calculation code COREBN-BWR

International Nuclear Information System (INIS)

Morimoto, Yuichi; Okumura, Keisuke

1992-05-01

In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

Numerical solution of matrix exponential in burn-up equation using mini-max polynomial approximation

International Nuclear Information System (INIS)

Kawamoto, Yosuke; Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

2015-01-01

Highlights: • We propose a new numerical solution of matrix exponential in burn-up depletion calculations. • The depletion calculation with extremely short half-lived nuclides can be done numerically stable with this method. • The computational time is shorter than the other conventional methods. - Abstract: Nuclear fuel burn-up depletion calculations are essential to compute the nuclear fuel composition transition. In the burn-up calculations, the matrix exponential method has been widely used. In the present paper, we propose a new numerical solution of the matrix exponential, a Mini-Max Polynomial Approximation (MMPA) method. This method is numerically stable for burn-up matrices with extremely short half-lived nuclides as the Chebyshev Rational Approximation Method (CRAM), and it has several advantages over CRAM. We also propose a multi-step calculation, a computational time reduction scheme of the MMPA method, which can perform simultaneously burn-up calculations with several time periods. The applicability of these methods has been theoretically and numerically proved for general burn-up matrices. The numerical verification has been performed, and it has been shown that these methods have high precision equivalent to CRAM

A Criticality Evaluation of the GBC-32 Dry Storage Cask in PWR Burnup Credit

Energy Technology Data Exchange (ETDEWEB)

Yoon, Hyoungju; Park, Kwangheon; Hong, Ser Gi [Kyung Hee Univ., Yongin (Korea, Republic of)

2015-05-15

The current criticality safety evaluation assumes the only unirradiated fresh fuels with the maximum enrichment in a dry storage cask (DSC) for conservatism without consideration of the depletion of fissile nuclides and the generation of neutron-absorbing fission products. However, the large conservatism leads to the significant increase of the storage casks required. Thus, the application of burnup credit which takes credit for the reduction of reactivity resulted from fuel depletion can increase the capacity in storage casks. On the other hand, the burnup credit application introduces lots of complexity into a criticality safety analysis such as the accurate estimation of the isotopic inventories and the burnup of UNFs and the validation of the criticality calculation. The criticality evaluation with an effect of burnup credit was performed for the DSC of GBC-32 by using SCALE 6.1/STARBUCS. keff values were calculated as a function of burnup and cooling time for four initial enrichments of 2, 3, 4, and 5 wt. % 235U. The values were calculated for the burnup range of 0 to 60,000 MWD/MTU, in increments of 10,000 MWD/MTU, and for five cooling times of 0, 5, 10, 20, and 40 years.

Regulatory status of burnup credit for storage and transport of spent fuel in Germany

International Nuclear Information System (INIS)

Neuber, J.C.; Schweer, H.H.; Johann, H.G.

2001-01-01

This paper describes the regulatory status of burnup credit applications to pond storage and dry-cask transport and storage of spent fuel in Germany. Burnup credit for wet storage of LWR fuel at nuclear power plants has to comply with the newly developed safety standard DIN 25471. This standard establishes the safety requirements for burnup credit criticality safety analysis of LWR fuel storage ponds and gives guidance on meeting these requirements. Licensing evaluations of dry transport systems are based on the application of the IAEA Safety Standards Series No.ST-1. However, because of the fact that burnup credit for dry-cask transport becomes more and more inevitable due to increasing initial enrichment of the fuel, and because of the increasing importance of dry-cask storage in Germany, the necessity of giving regulatory guidance on applying burnup credit to dry-cask transport and storage is seen. (author)

Analysis of collective life-cycle dose for burnup credit shipping casks

International Nuclear Information System (INIS)

Brentlinger, L.A.; Peterson, R.W.; Hofmann, P.L.

1989-01-01

In 1987, several studies were conducted by Sandia National Laboratories (SNL) to investigate the feasibility of and the incentive to justify the consideration of spent fuel histories in the design of spent fuel shipping casks. Taking credit for reduction in fissile content of fuel elements resulting from burnup credit is not current practice in the design and certification of shipping casks. The general argument can be made, however, that if this were done cask capacities could be increased over the current shipping cask designs which do not take the benefit of such burnup credit. This paper deals specifically with the question of occupational and public dose reduction via the use of a series of postulated burnup-credit cask designs

Numerical analysis and simulation of behavior of high burn-up PWR fuel pulse-irradiated in reactivity-initiated accident conditions

International Nuclear Information System (INIS)

Suzuki, M.; Sugiyama, T.; Udagawa, Y.; Nagase, F.; Fuketa, T.

2010-01-01

The four cases of the NSRR experiments, consisting of two room temperature tests and two high temperature tests, using high burn-up PWR fuel rods are analyzed by using the RANNS code to discuss the fuel behavior in hypothetical pulse-irradiation conditions, and the results are compared with metallography observations of ruptured claddings. The cladding rupture occurred by a shear sliding which starts from the tip of incipient crack generated in the hydride dense layer. The analyses reveal that the onset of shear sliding leading to cladding rupture can be closely associated with the stress intensity factor KI at the crack tip and local plastic strain evolution around the tip as well, and that these two factors depend also on the temperature of cladding. Simulation calculations on the basis of experimental conditions reveals that the cladding stress is dependent on the height and half-width of pulse power, and for the same integral enthalpy of pulse a larger half-width mitigates the severity of transient and decreases KI to allow plastic strain by temperature rise, thus failure possibility would be markedly decreased

Calculation of effect of burnup history on spent fuel reactivity based on CASMO5

International Nuclear Information System (INIS)

Li Xiaobo; Xia Zhaodong; Zhu Qingfu

2015-01-01

Based on the burnup credit of actinides + fission products (APU-2) which are usually considered in spent fuel package, the effect of power density and operating history on k_∞ was studied. All the burnup calculations are based on the two-dimensional fuel assembly burnup program CASMO5. The results show that taking the core average power density of specified power plus a bounding margin of 0.0023 to k_∞, and taking the operating history of specified power without shutdown during cycle and between cycles plus a bounding margin of 0.0045 to k_∞ can meet the bounding principle of burnup credit. (authors)

Models for fuel rod behaviour at high burnup

Energy Technology Data Exchange (ETDEWEB)

Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

2004-12-01

This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

Investigation of Burnup Credit Issues in BWR Fuel

International Nuclear Information System (INIS)

Broadhead, B.L.; DeHart, M.D.

1999-01-01

Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of 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, followed by criticality calculations to assess the value of keff for 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 used to characterize spent fuel. Most effort in the United States this decade has focused on burnup issues related to pressurized-water reactors. However, requirements for the permanent disposal of fuel from boiling-water reactors has necessitated development of methods for prediction of spent fuel contents for such fuels. Concomitant with such analyses, validation is also necessary. This paper provides a summary of initial efforts at the Oak Ridge National Laboratory to better understand and validate spent fuel analyses for boiling-water-reactor fuel

Development of continuous energy Monte Carlo burn-up calculation code MVP-BURN

International Nuclear Information System (INIS)

Okumura, Keisuke; Nakagawa, Masayuki; Sasaki, Makoto

2001-01-01

Burn-up calculations based on the continuous energy Monte Carlo method became possible by development of MVP-BURN. To confirm the reliably of MVP-BURN, it was applied to the two numerical benchmark problems; cell burn-up calculations for High Conversion LWR lattice and BWR lattice with burnable poison rods. Major burn-up parameters have shown good agreements with the results obtained by a deterministic code (SRAC95). Furthermore, spent fuel composition calculated by MVP-BURN was compared with measured one. Atomic number densities of major actinides at 34 GWd/t could be predicted within 10% accuracy. (author)

Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance

International Nuclear Information System (INIS)

Wagner, John C.; Parks, Cecil V.; Mueller, Don; Gauld, Ian C.

2010-01-01

Taking credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transport, disposal, and reprocessing of spent nuclear fuel (SNF) while maintaining sufficient subcritical margin to establish an adequate safety basis. Consequently, there continues to be considerable interest in the United States (U.S.), as well as internationally, in the increased use of burnup credit in SNF operations, particularly related to storage, transport, and disposal of commercial SNF. This interest has motivated numerous technical studies related to the application of burnup credit, both domestically and internationally, as well as the design of SNF storage, transport and disposal systems that rely on burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission (NRC) initiated a burnup credit research program in 1999, with support from the Oak Ridge National Laboratory (ORNL), to develop regulatory guidance and the supporting technical bases for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Although this NRC research program has not been continuous since its inception, considerable progress has been achieved in many key areas in terms of increased understanding of relevant phenomena and issues, availability of relevant information and data, and subsequently updated regulatory guidance for expanded use of burnup credit. This paper reviews technical studies performed by ORNL for the U.S. NRC burnup credit research program. Examples of topics include reactivity effects associated with reactor operating characteristics, fuel assembly characteristics, burnable absorbers, control rods, spatial burnup distributions, cooling time, and assembly misloading; methods and data for validation of isotopic composition predictions; methods and data for validation of criticality calculations; and

Characterisation of high-burnup LWR fuel rods through gamma tomography

International Nuclear Information System (INIS)

Caruso, S.

2007-01-01

Current fuel management strategies for light water reactors (LWRs), in countries with high back-end costs, progressively extend the discharge burnup at the expense of increasing the 235 U enrichment of the fresh UO 2 fuel loaded. In this perspective, standard non-destructive assay techniques, which are very attractive because they are fast, cheap, and preserve the fuel integrity, in contrast to destructive approaches, require further validation when burnup values become higher than 50 GWd/t. This doctoral work has been devoted to the development and optimisation of non-destructive assay techniques based on gamma-ray emissions from irradiated fuel. It represents an important extension of the unique, high-burnup related database, generated in the framework of the LWR PROTEUS Phase II experiments. A novel tomographic measurement station has been designed and developed for the investigation of irradiated fuel rod segments. A unique feature of the station is that it allows both gamma-ray transmission and emission computerised tomography to be performed on single fuel rods. Four burnt UO 2 fuel rod segments of 400 mm length have been investigated, two with very high (52 GWd/t and 71 GWd/t) and two with ultra-high (91 GWd/t and 126 GWd/t) burnup. Several research areas have been addressed, as described below. The application of transmission tomography to spent fuel rods has been a major task, because of difficulties of implementation and the uniqueness of the experiments. The main achievements, in this context, have been the determination of fuel rod average material density (a linear relationship between density and burnup was established), fuel rod linear attenuation coefficient distribution (for use in emission tomography), and fuel rod material density distribution. The non-destructive technique of emission computerised tomography (CT) has been applied to the very high and ultra-high burnup fuel rod samples for determining their within-rod distributions of caesium and

Preparation of computer codes for analyzing sensitivity coefficients of burnup characteristics (2) (Contract research, translated document)

International Nuclear Information System (INIS)

Hanaki, Hiroshi; Sanda, Toshio; Ohashi, Masahisa

2008-10-01

To develop nuclear design of LMFBR cores, they are important subjects of research and development to improve the accuracy in nuclear design of large LMFBR cores and to design highly efficient core more rationally. The adjusted nuclear cross-sections library has been made by being reflected the result of critical experiment of the JUPITER, etc. effectively as much as possible. And the distinct improvement of the accuracy in nuclear design of large LMFBR cores has been achieved. In the design of large LMFBR cores, however, it is important to accurately estimate not only nuclear characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. Therefore, it is thought to improve the prediction accuracy for burnup characteristics using many burnup data of 'Joyo' effectively. It is thought the best way to adjust cross sections using sensitivity coefficients of burnup characteristics to utilize burnup data of 'Joyo'. It is able to know the accuracy quantitatively for burnup characteristics of large LMFBR by analyzing the sensitivity coefficients. Therefore in this work computer codes for analyzing sensitivity coefficients of burnup characteristics had been prepared since 1992. In 1992 cross-section adjustment was done by using the data of 'Joyo' and the effect was studied. In this year the adequacy of the codes was studied with a view of applying of design of large LMFBR cores. The results are as follows: (1) The computer codes which could analyze sensitivity coefficients of burnup characteristics taking into consideration plural cycles and refueling were prepared, therefore it came of be able to adjust cross sections using burnup data and to estimate the accuracy for design of large LMFBR cores. The characteristics are not only burnup reactivity loss, breeding ratio but also number density, criticality, reactivity worth, reaction rate ratio, and reaction rate

Surface modification of 17-4PH stainless steel by DC plasma nitriding and titanium nitride film duplex treatment

International Nuclear Information System (INIS)

Qi, F.; Leng, Y.X.; Huang, N.; Bai, B.; Zhang, P.Ch.

2007-01-01

17-4PH stainless steel was modified by direct current (DC) plasma nitriding and titanium nitride film duplex treatment in this study. The microstructure, wear resistance and corrosion resistance were characterized by X-ray diffraction (XRD), pin-on-disk tribological test and polarization experiment. The results revealed that the DC plasma nitriding pretreatment was in favor of improving properties of titanium nitride film. The corrosion resistance and wear resistance of duplex treatment specimen was more superior to that of only coated titanium nitride film

First steps towards modelling high burnup effect in UO{sub 2} fuel

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

High burnup initiates a process that can lead to major microstructural changes near the edge of the fuel: formation of subgrains, the loss of matrix fission gas and an increase in porosity. A consequence of this, is a decrease of thermal conductivity near the edge of the fuel which may be major implications for the performance of LWR fuels at higher burnup. The mechanism for the changes in grain structure, the apparent depletion of Xe and increase in porosity is associated with the high fission density at the fuel periphery. This is in turn due to the preferential capture of epithermal neutrons in the resonances of {sup 238}U. The new model TUBRNP predicts the radial burnup profile as a function of time together with the radial profile of plutonium. The model has been validated with data from LWR UO{sub 2} fuels with enrichments in the range 2 to 8.25% and burnups between 21 to 75 Gwd/t. It has been reported that at high burnup EPMA measures a sharp decrease in the concentration of Xe near the fuel surface. This loss of Xe is interpreted as a signal that the gas has been swept out of the original grains into pores: this ``missing`` Xe has been measured by XRF. It has been noted experimentally that the restructuring (Xe depletion and changes in grain structure) have an onset threshold local burnup in the region of 70 to 80 GWd/t: a specific value was taken for use in the model. For a given fuel TUBRNP predicts the local burnup profile, and the depth corresponding to the threshold value is taken to be the thickness of the Xe depleted region. The theoretical predictions have been compared with experimental data. The results are presented and should be seen as a first step in the development of a more detailed model of this phenomenon. (author). 22 refs, 9 figs, 2 tabs.

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

International Nuclear Information System (INIS)

Lebrun, A.; Bignan, G.

2001-01-01

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

Metal surface nitriding by laser induced plasma

Science.gov (United States)

Thomann, A. L.; Boulmer-Leborgne, C.; Andreazza-Vignolle, C.; Andreazza, P.; Hermann, J.; Blondiaux, G.

1996-10-01

We study a nitriding technique of metals by means of laser induced plasma. The synthesized layers are composed of a nitrogen concentration gradient over several μm depth, and are expected to be useful for tribological applications with no adhesion problem. The nitriding method is tested on the synthesis of titanium nitride which is a well-known compound, obtained at present by many deposition and diffusion techniques. In the method of interest, a laser beam is focused on a titanium target in a nitrogen atmosphere, leading to the creation of a plasma over the metal surface. In order to understand the layer formation, it is necessary to characterize the plasma as well as the surface that it has been in contact with. Progressive nitrogen incorporation in the titanium lattice and TiN synthesis are studied by characterizing samples prepared with increasing laser shot number (100-4000). The role of the laser wavelength is also inspected by comparing layers obtained with two kinds of pulsed lasers: a transversal-excited-atmospheric-pressure-CO2 laser (λ=10.6 μm) and a XeCl excimer laser (λ=308 nm). Simulations of the target temperature rise under laser irradiation are performed, which evidence differences in the initial laser/material interaction (material heated thickness, heating time duration, etc.) depending on the laser features (wavelength and pulse time duration). Results from plasma characterization also point out that the plasma composition and propagation mode depend on the laser wavelength. Correlation of these results with those obtained from layer analyses shows at first the important role played by the plasma in the nitrogen incorporation. Its presence is necessary and allows N2 dissociation and a better energy coupling with the target. Second, it appears that the nitrogen diffusion governs the nitriding process. The study of the metal nitriding efficiency, depending on the laser used, allows us to explain the differences observed in the layer features

Reconstruction of pin burnup characteristics from nodal calculations in hexagonal geometry

International Nuclear Information System (INIS)

Yang, W.S.; Finck, P.J.; Khalil, H.S.

1990-01-01

A reconstruction method has been developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D/REBUS-3 code system. Intra-modal distributions of group fluxes, nuclide densities, power density, burnup, and fluence are efficiently computed using polynomial shapes constrained to satisfy nodal information. The accuracy of the method has been tested by performing several numerical benchmark calculations and by comparing predicted local burnups to values measured for experimental assemblies in EBR-11. The results indicate that the reconstruction methods are quite accurate, yielding maximum errors in power and nuclide densities that are less than 2% for driver assemblies and typically less than 5% for blanket assemblies. 14 refs., 2 figs., 5 tabs

Device for measuring a burnup degree

International Nuclear Information System (INIS)

Ito, Toshiaki; Goto, Seiichiro

1979-01-01

Purpose: To measure the burnup degree at high efficiency and accuracy. Constitution: The outer metal wall of fuel assemblies is heated under gamma radiation with long half life gamma rays in inverse proportion to the burnup degree and issues infrared radiation in proportion to the intensity of the gamma rays. An image pick-up tube is opposed to one surface of the fuel assemblies to detect the radiated infrared rays. Since the output signal from the pick-up tube is subjected to the absorptive damping by the distance between the pick-up tube and the fuel assembly, as well as water filled in the gap therebetween, it is corrected through a main amplifier comprising a signal correction circuit composed of a characteristic section inverse to the absorption property and a characteristic section inverse to the square of the distance. The corrected output signal is displayed on a display unit such as CRT or recorded in a film or a magnetic tape. (Furukawa, Y.)

Nitriding of high speed steel

International Nuclear Information System (INIS)

Doyle, E.D.; Pagon, A.M.; Hubbard, P.; Dowey, S.J.; Pilkington, A.; McCulloch, D.G.; Latham, K.; DuPlessis, J.

2010-01-01

Current practice when nitriding HSS cutting tools is to avoid embrittlement of the cutting edge by limiting the depth of the diffusion zone. This is accomplished by reducing the nitriding time and temperature and eliminating any compound layer formation. However, in many applications there is an argument for generating a compound layer with beneficial tribological properties. In this investigation results are presented of a metallographic, XRD and XPS analysis of nitrided surface layers generated using active screen plasma nitriding and reactive vapour deposition using cathodic arc. These results are discussed in the context of built up edge formation observed while machining inside a scanning electron microscope. (author)

Isotopic validation for PWR actinide-only burnup credit using Yankee Rowe data

International Nuclear Information System (INIS)

1997-11-01

Safety analyses of criticality control systems for transportation packages include an assumption that the spent nuclear fuel (SNF) loaded into the package is fresh or unirradiated. In other words, the spent fuel is assumed to have its original, as-manufactured U-235 isotopic content. The ''fresh fuel'' assumption is very conservative since the potential reactivity of the nuclear fuel is substantially reduced after being irradiated in the reactor core. The concept of taking credit for this reduction in nuclear fuel reactivity due to burnup of the fuel, instead of using the fresh fuel assumption in the criticality safety analysis, is referred to as ''Burnup Credit.'' Burnup credit uses the actual physical composition of the fuel and accounts for the net reduction of fissile material and the buildup of neutron absorbers in the fuel as it is irradiated. Neutron absorbers include actinides and other isotopes generated as a result of the fission process. Using only the change in actinide isotopes in the burnup credit criticality analysis is referred to as ''Actinide-Only Burnup Credit.'' The use of burnup credit in the design of criticality control systems enables more spent fuel to be placed in a package. Increased package capacity results in a reduced number of storage, shipping and disposal containers for a given number of SNF assemblies. Fewer shipments result in a lower risk of accidents associated with the handling and transportation of spent fuel, thus reducing both radiological and nonradiological risk to the public. This paper describes the modeling and the results of comparison between measured and calculated isotopic inventories for a selected number of samples taken from a Yankee Rowe spent fuel assembly

Modelling the evolution of composition-and stress-depth profiles in austenitic stainless steels during low-temperature nitriding

DEFF Research Database (Denmark)

Jespersen, Freja Nygaard; Hattel, Jesper Henri; Somers, Marcel A. J.

2016-01-01

. In the present paper solid mechanics was combined with thermodynamics and diffusion kinetics to simulate the evolution of composition-depth and stress-depth profiles resulting from nitriding. The model takes into account a composition-dependent diffusion coefficient of nitrogen in expanded austenite, short range......Nitriding of stainless steel causes a surface zone of expanded austenite, which improves the wear resistance of the stainless steel while preserving the stainless behaviour. During nitriding huge residual stresses are introduced in the treated zone, arising from the volume expansion...... that accompanies the dissolution of high nitrogen contents in expanded austenite. An intriguing phenomenon during low-temperature nitriding is that the residual stresses evoked by dissolution of nitrogen in the solid state, affect the thermodynamics and the diffusion kinetics of nitrogen dissolution...

End effects in the criticality analysis of burnup credit casks

International Nuclear Information System (INIS)

Brady, M.C.; Parks, C.V.

1990-01-01

A study to evaluate the effect of axially dependent burnup on k eff has been performed as part of an effort to qualify procedures to be used in establishing burnup credit in shipping cask design and certification. This study was performed using a generic 31-element modular cast-iron cask (wall thickness 33.1 cm) with a 1-cm-thick borated stainless-steel basket for reactivity control. Fuel isotopics used here are those of the 17 x 17 Westinghouse assemblies from the North Anna Unit 1 reactor. Virginia Power (VP) provided detailed spatial isotopics for the fuel assemblies in-core at beginning-of-cycle 5 (BOC-5) as generated from their PDQ analyses. Twenty-two axial planes were defined in the original VP data. The isotopics used in this study were for a 3.41 initial wt % 235 U and an average burnup of 31.5 GWd/MTU

Numerical solution of stiff burnup equation with short half lived nuclides by the Krylov subspace method

International Nuclear Information System (INIS)

Yamamoto, Akio; Tatsumi, Masahiro; Sugimura, Naoki

2007-01-01

The Krylov subspace method is applied to solve nuclide burnup equations used for lattice physics calculations. The Krylov method is an efficient approach for solving ordinary differential equations with stiff nature such as the nuclide burnup with short lived nuclides. Some mathematical fundamentals of the Krylov subspace method and its application to burnup equations are discussed. Verification calculations are carried out in a PWR pin-cell geometry with UO 2 fuel. A detailed burnup chain that includes 193 fission products and 28 heavy nuclides is used in the verification calculations. Shortest half life found in the present burnup chain is approximately 30 s ( 106 Rh). Therefore, conventional methods (e.g., the Taylor series expansion with scaling and squaring) tend to require longer computation time due to numerical stiffness. Comparison with other numerical methods (e.g., the 4-th order Runge-Kutta-Gill) reveals that the Krylov subspace method can provide accurate solution for a detailed burnup chain used in the present study with short computation time. (author)

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)

Solvothermal synthesis: a new route for preparing nitrides

CERN Document Server

Demazeau, G; Denis, A; Largeteau, A

2002-01-01

Solvothermal synthesis appears to be an interesting route for preparing nitrides such as gallium nitride and aluminium nitride, using ammonia as solvent. A nitriding additive is used to perform the reaction and, in the case of gallium nitride, is encapsulated by melt gallium. The syntheses are performed in the temperature range 400-800 deg. C and in the pressure range 100-200 MPa. The synthesized powders are characterized by x-ray diffraction and scanning electron microscopy. Finely divided gallium nitride GaN and aluminium nitride AlN, both with wurtzite-type structure, can be obtained by this route.

Validation of SCALE-4 for burnup credit applications

International Nuclear Information System (INIS)

Bowman, S.M.; DeHart, M.D.; Parks, C.V.

1995-01-01

In the past, a criticality analysis of PWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fuel burnup in the design of casks that are used in the transport of spent light water reactor fuel to a repository, the increase in payload can lead to a significant reduction in the cost of transport and a potential reduction in the risk to the public. A portion of the work has been performed at ORNL in support of the US DOE efforts to demonstrate a validation approach for criticality safety methods to be used in burnup credit cask design. To date, the SCALE code system developed at ORNL has been the primary computational tool used by DOE to investigate technical issues related to burnup credit. The ANSI/ANS-8.1 criticality safety standard requires validation and benchmarking of the calculational methods used in evaluating criticality safety limits for applications outside reactors by correlation against critical experiments that are applicable. Numerous critical experiments for fresh PWR-type fuel in storage and transport configurations exist and can be used as part of a validation database. However, there are no critical experiments with burned PWR-type fuel in storage and transport configurations. As an alternative, commercial reactors offer an excellent source of measured critical configurations. The results reported demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of k eff very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond these results, additional work in the determination of biases and uncertainties is necessary prior to use in burnup credit applications

Preparation of aluminum nitride-silicon carbide nanocomposite powder by the nitridation of aluminum silicon carbide

NARCIS (Netherlands)

Itatani, K.; Tsukamoto, R.; Delsing, A.C.A.; Hintzen, H.T.J.M.; Okada, I.

2002-01-01

Aluminum nitride (AlN)-silicon carbide (SiC) nanocomposite powders were prepared by the nitridation of aluminum-silicon carbide (Al4SiC4) with the specific surface area of 15.5 m2·g-1. The powders nitrided at and above 1400°C for 3 h contained the 2H-phases which consisted of AlN-rich and SiC-rich

Burnup measurements of leader fuel elements

International Nuclear Information System (INIS)

Henriquez, C; Navarro, G; Pereda, C

2000-01-01

Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

International Nuclear Information System (INIS)

DeHart, M.D.

1993-01-01

Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are 149 Sm, 151 Sm, and 155 Gd

IFPE/TRIBULATION R1, Fuel Rod Behaviour at High Burnup

International Nuclear Information System (INIS)

Turnbull, J.A.

2002-01-01

Description: The TRIBULATION (Tests Relative to High Burnup Limitations Arising Normally in LWRs) International Programme started in July 1980 and was organized jointly by BelgoNucleaire and the Nuclear Energy Centre at Mol (CEN/SCK) with the co-sponsorship of 14 participating organizations. The objectives of the programme were twofold. It was primarily a demonstration programme aimed at assessing the fuel rod behaviour at high burn-up, when an earlier transient had occurred in the power plant. The second objective was to investigate the behaviour of different fuel rod designs and manufacturers when subjected to a steady state irradiation history to high burn-up. The first objective was met by irradiating fuel rods under steady state conditions in the BR3 reactor and under transient conditions in BR2. The effect of the transient was determined by comparing data from 4 identical rods tested as follows: i) BR3 irradiation followed by PIE; ii) BR3 irradiation followed by BR2 transient then PIE; iii) BR3 irradiation followed by BR2 transient and re-irradiated in BR3 before PIE; iv) BR3 irradiation and continued BR3 irradiation to maximum burn-up before PIE. The Database contains data from 19 cases using rods fabricated by BelgoNucleaire (BN) (11) and Brown Boveri Reactor GmbH (BBR) (8)

Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

Energy Technology Data Exchange (ETDEWEB)

Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L. [Moscow Engineering Physics Institute (State University) (Russian Federation); Saito, M. [Tokyo Institute of Technology (Japan)

2003-07-01

Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, {sup 237}Np, {sup 238}Pu, {sup 231}Pa, {sup 232}U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations.

Deuterides of light elements: low-temperature thermonuclear burn-up and applications to thermonuclear fusion problems

International Nuclear Information System (INIS)

Frolov, A.M.; Smith, V.H.; Smith, G.T.

2002-01-01

Thermonuclear burn-up and thermonuclear applications are discussed for a number of deuterides and DT hydrides of light elements. These deuterides and corresponding DT hydrides are often used as thermonuclear fuels or components of such fuels. In fact, only for these substances thermonuclear energy gain exceeds (at some densities and temperatures) the bremsstrahlung loss and other high-temperature losses, i.e., thermonuclear burn-up is possible. Herein, thermonuclear burn-up in these deuterides and DT hydrides is considered in detail. In particular, a simple method is proposed to determine the critical values of the burn-up parameter x c for these substances and their mixtures at different temperatures and densities. The results for equimolar DT mixtures coincide quite well with the results of previous calculations. Also, the natural or Z limit is determined for low-temperature thermonuclear burn-up in the deuterides of light elements. (author)

Apatite formability of boron nitride nanotubes

International Nuclear Information System (INIS)

Lahiri, Debrupa; Keshri, Anup K; Agarwal, Arvind; Singh, Virendra; Seal, Sudipta

2011-01-01

This study investigates the ability of boron nitride nanotubes (BNNTs) to induce apatite formation in a simulated body fluid environment for a period of 7, 14 and 28 days. BNNTs, when soaked in the simulated body fluid, are found to induce hydroxyapatite (HA) precipitation on their surface. The precipitation process has an initial incubation period of ∼ 4.6 days. The amount of HA precipitate increases gradually with the soaking time. High resolution TEM results indicated a hexagonal crystal structure of HA needles. No specific crystallographic orientation relationship is observed between BNNT and HA, which is due to the presence of a thin amorphous HA layer on the BNNT surface that disturbs a definite orientation relationship.

Nuclear fuel burn-up economy; Ekonomija izgaranja nuklearnog goriva

Energy Technology Data Exchange (ETDEWEB)

Matausek, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1984-07-01

In the period 1981-1985, for the needs of Utility Organization, Beograd, and with the support of the Scientific Council of SR Srbija, work has been performed on the study entitled 'Nuclear Fuel Burn-up Economy'. The forst [phase, completed during the year 1983 comprised: comparative analysis of commercial NPP from the standpoint of nuclear fuel requirements; development of methods for fuel burn-up analysis; specification of elements concerning the nuclear fuel for the tender documentation. The present paper gives the short description of the purpose, content and results achieved in the up-to-now work on the study. (author)

Isotopic and criticality validation for actinide-only burnup credit

International Nuclear Information System (INIS)

Fuentes, E.; Lancaster, D.; Rahimi, M.

1997-01-01

The techniques used for actinide-only burnup credit isotopic validation and criticality validation are presented and discussed. Trending analyses have been incorporated into both methodologies, requiring biases and uncertainties to be treated as a function of the trending parameters. The isotopic validation is demonstrated using the SAS2H module of SCALE 4.2, with the 27BURNUPLIB cross section library; correction factors are presented for each of the actinides in the burnup credit methodology. For the criticality validation, the demonstration is performed with the CSAS module of SCALE 4.2 and the 27BURNUPLIB, resulting in a validated upper safety limit

Power ramp tests of high burnup BWR segment rods

International Nuclear Information System (INIS)

Hayashi, H.; Etoh, Y.; Tsukuda, Y.; Shimada, S.; Sakurai, H.

2002-01-01

Lead use assemblies (LUAs) of high burnup 8x8 fuel design for Japanese BWRs were irradiated up to 5 cycles in Fukushima Daini Nuclear Power Station No. 2 Unit. Segment rods were installed in LUAs and used for power ramp tests in Japanese Material Test Reactor (JMTR). Post irradiation examinations (PIEs) of segment rods were carried out at Nippon Nuclear Fuel Development Co., Ltd. before and after ramp tests. Maximum linear heat rates of LUAs were kept above 300 W/cm in the first cycle, above 250 W/cm in the second and third cycles and decreased to 200 W/cm in the fourth cycle and 80 W/cm in the fifth cycle. The integrity of high burnup 8x8 fuel was confirmed up to the bundle burnup of 48 GWd/t after 5 cycles of irradiation. Systematic and high quality data were collected through detailed PIEs. The main results are as follows. The oxide on the outer surface of cladding tubes was uniform and its thickness was less than 20 micro-meter after 5 cycles of irradiation and was almost independent of burnup. Hydrogen contents in cladding tubes were less than 150 ppm after 5 cycles of irradiation, although hydrogen contents increased during the fourth and fifth irradiation cycles. Mechanical properties of cladding tubes were on the extrapolated line of previous data up to 5 cycles of irradiation. Fission gas release rates were in the low level (mainly less than 6%) up to 5 cycles of irradiation due to the design to decrease pellet temperature. Pellet-cladding bonding layers were observed after the third cycle and almost full bonding was observed after the fifth cycle. Pellet volume increased with burnup in proportion to solid swelling rate up to the forth cycle. After the fifth cycle, slightly higher pellet swelling was confirmed. Power ramp tests were carried out and satisfactory performance of Zr-lined cladding tube was confirmed up to 60 GWd/t (segment average burnup). One segment rod irradiated for 3 cycles failed by a single step ramp test at terminal ramp power of 614 W

Microstructural characterization of an AISI-SAE 4140 steel without nitridation and nitrided

International Nuclear Information System (INIS)

Medina F, A.; Naquid G, C.

2000-01-01

It was micro structurally characterized an AISI-SAE 4140 steel before and after of nitridation through the nitridation process by plasma post-unloading microwaves through Optical microscopy (OM), Scanning electron microscopy (SEM) by means of secondary electrons and retrodispersed, X-ray diffraction (XRD), Energy dispersion spectra (EDS) and mapping of elements. (Author)

Total surface area change of Uranium dioxide fuel in function of burn-up and its impact on fission gas release during neutron irradiation for small, intermediate and high burn-up

International Nuclear Information System (INIS)

Szuta, M.

2011-01-01

In the early published papers it was observed that the fractional fission gas release from the specimen have a tendency to increase with the total surface area of the specimen - a fairy linear relationship was indicated. Moreover it was observed that the increase of total surface area during irradiation occurs in the result of connection the closed porosity with the open porosity what in turn causes the increase of fission gas release. These observations let us surmise that the process of knock-out release is the most significant process of fission gas release since its quantity is proportional to the total surface area. Review of the experiments related to the increase of total surface area in function of burn-up is presented in the paper. For very high burn-up the process of grain sub-division (polygonization) occurs under condition that the temperature of irradiated fuel lies below the temperature of grain re-crystallization. Simultaneously with the process of polygonization, the increase in local porosity and the decrease in local density in function of burn-up occurs, which leads to the increase of total surface area. It is suggested that the same processes take place in the transformed fuel as in the original fuel, with the difference that the total surface area is so big that the whole fuel can be treated as that affected by the knock-out process. This leads to explanation of the experimental data that for very high burn-up (>120 MWd/kgU) the concentration of xenon is constant. An explanation of the grain subdivision process in function of burn-up in the 'athermal' rim region in terms of total surface area, initial grain size and knock-out release is undertaken. Correlation of the threshold burn-up, the local fission gas concentration, local total surface area, initial and local grain size and burn-up in the rim region is expected. (author)

Validation issues for depletion and criticality analysis in burnup credit

International Nuclear Information System (INIS)

Parks, C.V.; Broadhead, B.L.; Dehart, M.D.; Gauld, I.C.

2001-01-01

This paper reviews validation issues associated with implementation of burnup credit in transport, dry storage, and disposal. The issues discussed are ones that have been identified by one or more constituents of the United States technical community (national laboratories, licensees, and regulators) that have been exploring the use of burnup credit. There is not necessarily agreement on the importance of the various issues, which sometimes is what creates the issue. The broad issues relate to the paucity of available experimental data (radiochemical assays and critical experiments) covering the full range and characteristics of spent nuclear fuel in away-from-reactor systems. The paper will also introduce recent efforts initiated at Oak Ridge National Laboratory (ORNL) to provide technical information that can help better assess the value of different experiments. The focus of the paper is on experience with validation issues related to use of burnup credit for transport and dry storage applications. (author)

Calculation study of the WWER-440 fuel performance for extended burnup

International Nuclear Information System (INIS)

Kujal, J.; Pazdera, F.; Barta, O.

1984-01-01

The results of preliminary calculational study of extended burnup cycling schemes impact on WWER-440 fuel performance are presented. Two high burnup schemes were proposed with three and four cycles, resp. Comparison was made with three cycle reference case. The thermal mechanical analysis was performed with PIN and RELA codes. The values of rod internal pressure, fuel centerline temperatures and fuel-cladding gap are expressed as function of power history. (author)

Precipitation of metal nitrides from chloride melts

International Nuclear Information System (INIS)

Slater, S.A.; Miller, W.E.; Willit, J.L.

1996-01-01

Precipitation of actinides, lanthanides, and fission products as nitrides from molten chloride melts is being investigated for use as a final cleanup step in treating radioactive salt wastes generated by electrometallurgical processing of spent nuclear fuel. The radioactive components (eg, fission products) need to be removed to reduce the volume of high-level waste that requires disposal. To extract the fission products from the salt, a nitride precipitation process is being developed. The salt waste is first contacted with a molten metal; after equilibrium is reached, a nitride is added to the metal phase. The insoluble nitrides can be recovered and converted to a borosilicate glass after air oxidation. For a bench-scale experimental setup, a crucible was designed to contact the salt and metal phases. Solubility tests were performed with candidate nitrides and metal nitrides for which there are no solubility data. Experiments were performed to assess feasibility of precipitation of metal nitrides from chloride melts

Burnup credit implementation in WWER spent fuel management systems: Status and future aspects

International Nuclear Information System (INIS)

Manolova, M.

1998-01-01

This paper describes the motivation for possible burnup credit implementation in WWER spent fuel management systems in Bulgaria. The activities being done are described, namely: the development and verification of a 3D few-group diffusion burnup model; the application of the KORIGEN code for evaluation of WWER fuel nuclear inventory during reactor core lifetime and after spent fuel discharge; using the SCALE modular system (PC Version 4.1) for criticality safety analyses of spent fuel storage facilities. Future plans involving such important tasks as validation and verification of computer systems and libraries for WWER burnup credit analysis are shown. (author)

A comparative study of MONTEBURNS and MCNPX 2.6.0 codes in ADS simulations

International Nuclear Information System (INIS)

Barros, Graiciany P.; Pereira, Claubia; Veloso, Maria A.F.; Velasquez, Carlos E.; Costa, Antonella L.

2013-01-01

The possible use of the MONTEBURNS and MCNPX 2.6.0 codes in Accelerator-driven systems (ADSs) simulations for fuel evolution description is discussed. ADSs are investigated for fuel breeding and long-lived fission product transmutation so simulations of fuel evolution have a great relevance. The burnup/depletion capability is present in both studied codes. MONTEBURNS code links Monte Carlo N-Particle Transport Code (MCNP) to the radioactive decay burnup code ORIGEN2, whereas MCNPX depletion/ burnup capability is a linked process involving steady-state flux calculations by MCNPX and nuclide depletion calculations by CINDER90. A lead-cooled accelerator-driven system fueled with thorium was simulated and the results obtained using MONTEBURNS code and the results from MCNPX 2.6.0 code were compared. The system criticality and the variation of the actinide inventory during the burnup were evaluated and the results indicate a similar behavior between the results of each code. (author)

Enhanced thermaly managed packaging for III-nitride light emitters

Science.gov (United States)

Kudsieh, Nicolas

In this Dissertation our work on `enhanced thermally managed packaging of high power semiconductor light sources for solid state lighting (SSL)' is presented. The motivation of this research and development is to design thermally high stable cost-efficient packaging of single and multi-chip arrays of III-nitrides wide bandgap semiconductor light sources through mathematical modeling and simulations. Major issues linked with this technology are device overheating which causes serious degradation in their illumination intensity and decrease in the lifetime. In the introduction the basics of III-nitrides WBG semiconductor light emitters are presented along with necessary thermal management of high power cingulated and multi-chip LEDs and laser diodes. This work starts at chip level followed by its extension to fully packaged lighting modules and devices. Different III-nitride structures of multi-quantum well InGaN/GaN and AlGaN/GaN based LEDs and LDs were analyzed using advanced modeling and simulation for different packaging designs and high thermal conductivity materials. Study started with basic surface mounted devices using conventional packaging strategies and was concluded with the latest thermal management of chip-on-plate (COP) method. Newly discovered high thermal conductivity materials have also been incorporated for this work. Our study also presents the new approach of 2D heat spreaders using such materials for SSL and micro LED array packaging. Most of the work has been presented in international conferences proceedings and peer review journals. Some of the latest work has also been submitted to well reputed international journals which are currently been reviewed for publication. .

Fuel element burnup measurements for the equilibrium LEU silicide RSG GAS (MPR-30) core under a new fuel management strategy

International Nuclear Information System (INIS)

Pinem, Surian; Liem, Peng Hong; Sembiring, Tagor Malem; Surbakti, Tukiran

2016-01-01

Highlights: • Burnup measurement of fuel elements comprising the new equilibrium LEU silicide core of RSG GAS. • The burnup measurement method is based on a linear relationship between reactivity and burnup. • Burnup verification was conducted using an in-house, in-core fuel management code BATAN-FUEL. • A good agreement between the measured and calculated burnup was confirmed. • The new fuel management strategy was confirmed and validated. - Abstract: After the equilibrium LEU silicide core of RSG GAS was achieved, there was a strong need to validate the new fuel management strategy by measuring burnup of fuel elements comprising the core. Since the regulatory body had a great concern on the safety limit of the silicide fuel element burnup, amongst the 35 burnt fuel elements we selected 22 fuel elements with high burnup classes i.e. from 20 to 53% loss of U-235 (declared values) for the present measurements. The burnup measurement method was based on a linear relationship between reactivity and burnup where the measurements were conducted under subcritical conditions using two fission counters of the reactor startup channel. The measurement results were compared with the declared burnup evaluated by an in-house in-core fuel management code, BATAN-FUEL. A good agreement between the measured burnup values and the calculated ones was found within 8% uncertainties. Possible major sources of differences were identified, i.e. large statistical errors (i.e. low fission counters’ count rates), variation of initial U-235 loading per fuel element and accuracy of control rod indicators. The measured burnup of the 22 fuel elements provided the confirmation of the core burnup distribution planned for the equilibrium LEU silicide core under the new fuel management strategy.

Effective Duration of Gas Nitriding Process on AISI 316L for the Formation of a Desired Thickness of Surface Nitrided Layer

Directory of Open Access Journals (Sweden)

Mahmoud Hassan R. S.

2014-07-01

Full Text Available High temperature gas nitriding performed on AISI 316L at the temperature of 1200°C. The microstructure of treated AISI 316L samples were observed to identify the formation of the microstructure of nitrided surface layer. The grain size of austenite tends to be enlarged when the nitriding time increases, but the austenite single phase structure is maintained even after the long-time solution nitriding. Using microhardness testing, the hardness values drop to the center of the samples. The increase in surface hardness is due to the high nitrogen concentration at or near the surface. At 245HV, the graph of the effective duration of nitriding process was plotted to achieve the maximum depth of nitrogen diffuse under the surface. Using Sigma Plot software best fit lines of the experimental result found and plotted to find out effective duration of nitriding equation as Y=1.9491(1-0.7947x, where Y is the thickness of nitrided layer below the surface and X is duration of nitriding process. Based on this equation, the duration of gas nitriding process can be estimated to produce desired thickness of nitrided layer.

OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

Energy Technology Data Exchange (ETDEWEB)

DeHart, M.D.

1993-01-01

Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are {sup 149}Sm, {sup 151}Sm, and {sup 155}Gd.

Modeling of WWER-440 Fuel Pin Behavior at Extended Burn-up

International Nuclear Information System (INIS)

El-Koliel, M.S.; Abou-Zaid, A.A.; El-Kafas, A.A.

2004-01-01

Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWER's as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased to 60 to 70 Mwd/kg U. The change in the fuel radial power distribution as a function of fuel burn up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. In this paper, the radial burn-up and fissile products distributions of WWER-440 UO 2 fuel pin were evaluated using MCNP 4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted fission gas release calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. a computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented

Separation of zirconium--hafnium by nitride precipitation

International Nuclear Information System (INIS)

Anderson, R.N.; Parlee, N.A.

1977-01-01

A method is described for the separation of a light reactive metal (e.g., zirconium) from a heavy reactive metal (e.g., hafnium) by forming insoluble nitrides of the metals in a molten metal solvent (e.g., copper) inert to nitrogen and having a suitable density for the light metal nitride to form a separate phase in the upper portion of the solvent and for the heavy metal nitride to form a separate phase in the lower portion of the solvent. Nitriding is performed by maintaining a nitrogen-containing atmosphere over the bath. The light and heavy metals may be an oxide mixture and carbothermically reduced to metal form in the same bath used for nitriding. The nitrides are then separately removed and decomposed to form the desired separate metals. 16 claims, 1 figure

Calculation of heat rating and burn-up for test fuel pins irradiated in DR 3

International Nuclear Information System (INIS)

Bagger, C.; Carlsen, H.; Hansen, K.

1980-01-01

A summary of the DR 3 reactor and HP1 rig design is given followed by a detailed description of the calculation procedure for obtaining linear heat rating and burn-up values of fuel pins irradiated in HP1 rigs. The calculations are carried out rather detailed, especially regarding features like end pellet contribution to power as a function of burn-up, gamma heat contributions, and evaluation of local values of heat rating and burn-up. Included in the report is also a description of the fast flux- and cladding temperature calculation techniques currently used. A good agreement between measured and calculated local burn-up values is found. This gives confidence to the detailed treatment of the data. (author)

Nitride fuels irradiation performance data base

International Nuclear Information System (INIS)

Brozak, D.E.; Thomas, J.K.; Peddicord, K.L.

1987-01-01

An irradiation performance data base for nitride fuels has been developed from an extensive literature search and review that emphasized uranium nitride, but also included performance data for mixed nitrides [(U,Pu)N] and carbonitrides [(U,Pu)C,N] to increase the quantity and depth of pin data available. This work represents a very extensive effort to systematically collect and organize irradiation data for nitride-based fuels. The data base has many potential applications. First, it can facilitate parametric studies of nitride-based fuels to be performed using a wide range of pin designs and operating conditions. This should aid in the identification of important parameters and design requirements for multimegawatt and SP-100 fuel systems. Secondly, the data base can be used to evaluate fuel performance models. For detailed studies, it can serve as a guide to selecting a small group of pin specimens for extensive characterization. Finally, the data base will serve as an easily accessible and expandable source of irradiation performance information for nitride fuels

UO2 fuel behaviour at rod burn-ups up to 105 MWd/kgHM. A review of 10 years of high burn-up examinations commissioned by AREVA NP

International Nuclear Information System (INIS)

Goll, W.; Hoffmann, P.B.; Hellwig, C.; Sauser, W.; Spino, J.; Walker, C.T.

2007-01-01

Irradiation experience gained on fuel rods with burn-ups greater than 60 MWd/kgHM irradiated in the Nuclear Power Plant Goesgen, Switzerland, is described. Emphasis is placed on the fuel behaviour, which has been analysed by hot cell examinations at the Institute for Transuranium Elements and the Paul-Scherrer-Institute. Above 60 MWd/kgHM, the so-called high burn-up structure (HBS) forms and the fission gas release increases with burn-up and rod power. Examinations performed in the outer region of the fuel revealed that most if not all of the fission gas created was retained in the HBS, even at 25% porosity. Furthermore, the HBS has a relatively low swelling rate, greatly increased plasticity, and its thermal conductivity is higher than expected from the porosity. The post-irradiation examinations showed that the HBS has no detrimental effects on the performance of stationary irradiated PWR fuel irradiated to the high burn-ups that can be achieved with 5 wt% U-235 enrichment. On the contrary, the HBS results in fuel performance that is generally better than it would have been if the HBS had not formed. (orig.)

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)

Simple process to fabricate nitride alloy powders

International Nuclear Information System (INIS)

Yang, Jae Ho; Kim, Dong-Joo; Kim, Keon Sik; Rhee, Young Woo; Oh, Jang-Soo; Kim, Jong Hun; Koo, Yang Hyun

2013-01-01

Uranium mono-nitride (UN) is considered as a fuel material [1] for accident-tolerant fuel to compensate for the loss of fissile fuel material caused by adopting a thickened cladding such as SiC composites. Uranium nitride powders can be fabricated by a carbothermic reduction of the oxide powders, or the nitriding of metal uranium. Among them, a direct nitriding process of metal is more attractive because it has advantages in the mass production of high-purity powders and the reusing of expensive 15 N 2 gas. However, since metal uranium is usually fabricated in the form of bulk ingots, it has a drawback in the fabrication of fine powders. The Korea Atomic Energy Research Institute (KAERI) has a centrifugal atomisation technique to fabricate uranium and uranium alloy powders. In this study, a simple reaction method was tested to fabricate nitride fuel powders directly from uranium metal alloy powders. Spherical powder and flake of uranium metal alloys were fabricated using a centrifugal atomisation method. The nitride powders were obtained by thermal treating the metal particles under nitrogen containing gas. The phase and morphology evolutions of powders were investigated during the nitriding process. A phase analysis of nitride powders was also part of the present work. KAERI has developed the centrifugal rotating disk atomisation process to fabricate spherical uranium metal alloy powders which are used as advanced fuel materials for research reactors. The rotating disk atomisation system involves the tasks of melting, atomising, and collecting. A nozzle in the bottom of melting crucible introduces melt at the center of a spinning disk. The centrifugal force carries the melt to the edge of the disk and throws the melt off the edge. Size and shape of droplets can be controlled by changing the nozzle size, the disk diameter and disk speed independently or simultaneously. By adjusting the processing parameters of the centrifugal atomiser, a spherical and flake shape

Challenges in the application of burn-up credit to the criticality safety of the THORP reprocessing plant

International Nuclear Information System (INIS)

Mayson, R.T.H.; Gunston, K.J.

1999-01-01

Since 1991 BNFL has made a significant investment in the development of the burn-up credit method and the application to its operations. It has recently demonstrated that using this method for the THORP dissolvers, it is possible to justify operating safety with reduced neutron poison concentrations and this has now been submitted to the regulators. The continued challenges the criticality safety community is facing are to show that we are not reducing safety levels because we are using burn-up credit. The burn-up credit method that has been developed can be summarized as follows. It consists of performing reactivity calculations for irradiated fuel using compositions generated by and inventory prediction code, generally in order to determine the limiting burn-up required for that fuel in a particular environment. In addition, it has always been envisaged that a confirmatory measurement of burn-up would be required to be made prior to certain operations such as the sharing of fuel into a dissolver. The burn-up credit method therefore relies upon three key components of inventory prediction, reactivity calculation code and the quantification and verification of burn-up. (J.P.N.)

Three dimensional Burn-up program parallelization using socket programming

International Nuclear Information System (INIS)

Haliyati R, Evi; Su'ud, Zaki

2002-01-01

A computer parallelization process was built with a purpose to decrease execution time of a physics program. In this case, a multi computer system was built to be used to analyze burn-up process of a nuclear reactor. This multi computer system was design need using a protocol communication among sockets, i.e. TCP/IP. This system consists of computer as a server and the rest as clients. The server has a main control to all its clients. The server also divides the reactor core geometrically to in parts in accordance with the number of clients, each computer including the server has a task to conduct burn-up analysis of 1/n part of the total reactor core measure. This burn-up analysis was conducted simultaneously and in a parallel way by all computers, so a faster program execution time was achieved close to 1/n times that of one computer. Then an analysis was carried out and states that in order to calculate the density of atoms in a reactor of 91 cm x 91 cm x 116 cm, the usage of a parallel system of 2 computers has the highest efficiency

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

Microhardness and microplasticity of zirconium nitride

International Nuclear Information System (INIS)

Neshpor, V.S.; Eron'yan, M.A.; Petrov, A.N.; Kravchik, A.E.

1978-01-01

To experimentally check the concentration dependence of microhardness of 4 group nitrides, microhardness of zirconium nitride compact samples was measured. The samples were obtained either by bulk saturation of zirconium iodide plates or by chemical precipitation from gas. As nitrogen content decreased within the limits of homogeneity of zirconium nitride samples where the concentration of admixed oxygen was low, the microhardness grew from 1500+-100 kg/mm 2 for ZrNsub(1.0) to 27000+-100 kg/mm 2 for ZrNsub(0.78). Microplasticity of zirconium nitride (resistance to fracture) decreased, as the concentration of nitrogen vacancies was growing

Surface analysis in steel nitrides by using Moessbauer spectroscopy

International Nuclear Information System (INIS)

Figueiredo, R.S. de.

1991-07-01

The formation of iron nitride layer at low temperatures, 600-700 K, by Moessbauer spectroscopy is studied. These layers were obtained basically through two different processes: ion nitriding and ammonia gas nitriding. A preliminary study about post-discharge nitriding was made using discharge in hollow cathode as well as microwave excitation. The assembly of these chambers is also described. The analysis of the nitrided samples was done by CEMS and CXMS, aided by optical microscopy, and the CEMS and CXMS detectors were constructed by ourselves. We also made a brief study about these detectors, testing as acetone as the mixture 80% He+10% C H 4 as detection gases for the use of CEMS. The surface analysis of the samples showed that in the ammonia gas process nitriding the nitrided layer starts by the superficial formation of an iron nitride rich nitrogen. By thermal evolution this nitride promotes the diffusion of nitrogen and the formation of other more stable nitrides. (author)

Consequences of the increase of burnup on the fuel

International Nuclear Information System (INIS)

Melin, P.; Lavoine, O.; Houdaille, B.

1986-04-01

The examinations carried out on the FRAGEMA fuel of EDF reactors show its good behavior in service. The results of research and development programs developed by EDF, FGA and the CEA show that this fuel can be irradiated up to a high burnup, and allow to point out the axies of research to improve still the performance of the product in a more and more soliciting environment (increase of power and burnup coupled with load following). Among the solutions considered, there are the design and fabrication adjustments (geometry, initial pressurization), more fundamental changes concerning fuel cans and fuel pellets, which need still research and development programs [fr

Tag gas burnup based on three-dimensional FTR analysis

International Nuclear Information System (INIS)

Kidman, R.B.

1976-01-01

Flux spectra from a three-dimensional diffusion theory analysis of the Fast Test Reactor (FTR) are used to predict gas tag ratio changes, as a function of exposure, for each FTR fuel and absorber subassembly plenum. These flux spectra are also used to predict Xe-125 equilibrium activities in absorber plena in order to assess the feasibility of using Xe-125 gamma rays to detect and distinguish control rod failures from fuel rod failures. Worst case tag burnup changes are used in conjunction with burnup and mass spectrometer uncertainties to establish the minimum spacing of tags which allows the tags to be unambiguously identified

High Burnup Fuel Behaviour under LOCA Conditions as Observed in Halden Reactor Experiments

International Nuclear Information System (INIS)

Kolstad, E.; Wiesenack, W.; Oberlander, B.; Tverberg, T.

2013-01-01

In the context of assessing the validity of safety criteria for loss of coolant accidents with high burnup fuel, the OECD Halden Reactor Project has implemented an integral in-pile LOCA test series. In this series, fuel fragmentation and relocation, axial gas communication in high burnup rods as affected by gap closure and fuel- clad bonding, and secondary cladding oxidation and hydriding are of major interest. In addition, the data are being used for code validation as well as model development and verification. So far, nine tests with irradiated fuel segments (burnup 40-92 MW.d.kg -1 ) from PWR, BWR and VVER commercial nuclear power plants have been carried out. The in-pile measurements and the PIE results show a good repeatability of the experiments. The paper describes the experimental setup as well as the principal features and main results of these tests. Fuel fragmentation and relocation have occurred to varying degrees in these tests. The paper compares the conditions leading to the presence or absence of fuel fragmentation, e.g., burnup and loss of constraint. Axial gas flow is an important driving force for clad ballooning, fuel relocation and fuel expulsion. The experiments have provided evidence that such gas flow can be impeded in high burnup fuel with a potential impact on the ballooning and fuel dispersal. Although the results of the Halden LOCA tests are, to some extent, amplified by conditions and features deliberately introduced into the test series, the fuel behaviour identified in the Halden tests has an impact on the safety assessment of high burnup fuel and should give rise to improvements of the predictive capabilities of LOCA modelling codes. (author)

EPRI/DOE High-Burnup Fuel Sister Rod Test Plan Simplification and Visualization

Energy Technology Data Exchange (ETDEWEB)

Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hanson, B. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Shimskey, R. W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Klymyshyn, N. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Webster, R. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, P. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); MacFarlan, P. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billone, Mike [Argonne National Lab. (ANL), Argonne, IL (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-15

The EPRI/DOE High-Burnup Confirmatory Data Project (herein called the “Demo”) is a multi-year, multi-entity test with the purpose of providing quantitative and qualitative data to show if high-burnup fuel mechanical properties change in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of common cladding alloys from the North Anna Nuclear Power Plant, loading them in an NRC-licensed TN-32B cask, drying them according to standard plant procedures, and then storing them on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the mechanical properties of the rods will be tested and analyzed.

Burn-Up Determination by High Resolution Gamma Spectrometry: Fission Product Migration Studies

Energy Technology Data Exchange (ETDEWEB)

Forsyth, R S; Blackadder, W H; Ronqvist, N

1967-04-15

The migration of solid fission products, in particular caesium and ruthenium, in high temperature oxide fuel can create a severe problem during the application of non-destructive burn-up methods employing gamma spectrometry, since caesium-137 is otherwise the most convenient long-lived burn-up monitor and ruthenium-106 can be used to distinguish between fissions in U-235 and Pu-239. As part of an experimental programme to develop burn-up methods, gamma scanning experiments have been performed on slices of irradiated UO{sub 2} pellets using a lithium-drifted germanium detector. The usefulness of the technique for migration studies has been demonstrated by comparing the fission product distribution curves across the specimen diameters with the microstructure of the specimens after polishing and etching.

Fission gas and iodine release measured up to 15 GWd/t UO2 burnup

International Nuclear Information System (INIS)

Appelhans, A.D.

1983-01-01

A summary is presented of the measured release of xenon, krypton and iodine up to 15 GWd/t UO 2 burnup for fuel centerline temperatures ranging from 950 to 1800 K, at average linear heat ratings of 15 to 35 kW/m. The IFA-430 is composed of four 1.28-m-long fuel rods containing 10% enriched UO 2 pellet fuel. Two of the fuel rods are connected, top and bottom, to a gas flow system that permits the fission gases released from the fuel pellets to be swept out of the rods during irradiation and measured via gamma spectrometry. The release/burnup increased significantly between 10 and 15 GWd/t burnup. Fuel temperature did not change. Increased releases were due to physical changes in the fuel-surface area. Changes appeared to be due to higher power operation and burnup

Approach to lithium burn-up effect in lithium ceramics

International Nuclear Information System (INIS)

Rasneur, B.

1994-01-01

The lithium burn-up in Li 2 ZrO 3 is simulated by removing lithium under Li 2 O form and trapping it in high specific surface area powder while heating during 15 days or 1 month at moderate temperature so that lithium mobility be large enough without causing any sintering neither of the specimens nor of the powder. In a first treatment at 775 deg C during 1 month. 30% of the lithium content could be removed inducing a lithium concentration gradient in the specimen and the formation of a lithium-free monoclinic ZrO 2 skin. Improvements led to similar results at 650 deg C and 600 deg C, the latter temperatures are closer to the operating temperature of the ceramic breeder blanket of a fusion reactor. (author) 4 refs.; 4 figs.; 1 tab

Leachability of nitrided ilmenite in hydrochloric acid

OpenAIRE

Swanepoel, J.J.; van Vuuren, D.S.; Heydenrych, M.

2011-01-01

Titanium nitride in upgraded nitrided ilmenite (bulk of iron removed) can selectively be chlorinated to produce titanium tetrachloride. Except for iron, most other components present during this low temperature (ca. 200°C) chlorination reaction will not react with chlorine. It is therefore necessary to remove as much iron as possible from the nitrided ilmenite. Hydrochloric acid leaching is a possible process route to remove metallic iron from nitrided ilmenite without excessive dissolution o...

Effect of residual stresses on fatigue strength of plasma nitrided 4140 steel

International Nuclear Information System (INIS)

Aghazadeh, J.; Amidi, M.R.

2004-01-01

Almost every method that has been presented to determine residual stress has some limitation and complexities. The aim of this work is to present a new, yet simple method so called strain indentation for measuring the residual stresses particularly in thin layers. In this method in addition to the precision measurements, components of residual stress at different directions may be determined. AISI 4140 steel specimens nitrided at 350 d ig C , 450 d ig C and 550 d ig C for 5 hours in the mixture of 75% nitrogen- 25% hydrogen gas. The, components of residual stress in the radials axial and hoop directions in the nitrided layer were determined considering the elastic strain recovery after removal of residual stress inducer(i.e. the nitrided layer). Fatigue strength of the nitrided specimens was obtained by plotting the S-N curves and fractographic studies carried out on the fracture surface of the specimens. The effect of residual stress on the stress pattern was simulated. The calculated residual stress components were in the range of 40-210 Mpa and the radial components of residual stress were more than the other two directions. Maximum fatigue strength improvement of up to 110% was observed in the plasma nitrided specimens at 550 d ig C and also 40% improvement in fatigue strength was detected by increasing the nitriding temperature from 350 d ig C to 550 d ig C . This was due to 100% increase in residual stress. Fatigue crack growth velocity in the hoop direction was more than that of radial direction. This seems to be due to higher radial residual stress component compared with the hoop stress component in the sub layer

Implementation of burnup credit in spent fuel management systems. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-08-01

The purpose of this Technical Committee Meeting was to explore the status of international activities related to the use of burnup credit for spent fuel applications. This was the second major meeting on the issues of burnup credit for spent fuel management systems held since the IAEA began to monitor the uses of burnup credit in spent fuel management systems in 1997. Burnup credit for wet and dry storage systems is needed in many Member States to allow for increased initial fuel enrichment, and to increase the storage capacity and thus to avoid the need for extensive modifications of the spent fuel management systems involved. This document contains 31 individual papers presented at the Meeting; each of the papers was indexed separately.

Implementation of burnup credit in spent fuel management systems. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2001-08-01

The purpose of this Technical Committee Meeting was to explore the status of international activities related to the use of burnup credit for spent fuel applications. This was the second major meeting on the issues of burnup credit for spent fuel management systems held since the IAEA began to monitor the uses of burnup credit in spent fuel management systems in 1997. Burnup credit for wet and dry storage systems is needed in many Member States to allow for increased initial fuel enrichment, and to increase the storage capacity and thus to avoid the need for extensive modifications of the spent fuel management systems involved. This document contains 31 individual papers presented at the Meeting; each of the papers was indexed separately

Work Function Characterization of Potassium-Intercalated, Boron Nitride Doped Graphitic Petals

Directory of Open Access Journals (Sweden)

Patrick T. McCarthy

2017-07-01

Full Text Available This paper reports on characterization techniques for electron emission from potassium-intercalated boron nitride-modified graphitic petals (GPs. Carbon-based materials offer potentially good performance in electron emission applications owing to high thermal stability and a wide range of nanostructures that increase emission current via field enhancement. Furthermore, potassium adsorption and intercalation of carbon-based nanoscale emitters decreases work functions from approximately 4.6 eV to as low as 2.0 eV. In this study, boron nitride modifications of GPs were performed. Hexagonal boron nitride is a planar structure akin to graphene and has demonstrated useful chemical and electrical properties when embedded in graphitic layers. Photoemission induced by simulated solar excitation was employed to characterize the emitter electron energy distributions, and changes in the electron emission characteristics with respect to temperature identified annealing temperature limits. After several heating cycles, a single stable emission peak with work function of 2.8 eV was present for the intercalated GP sample up to 1,000 K. Up to 600 K, the potassium-intercalated boron nitride modified sample exhibited improved retention of potassium in the form of multiple emission peaks (1.8, 2.5, and 3.3 eV resulting in a large net electron emission relative to the unmodified graphitic sample. However, upon further heating to 1,000 K, the unmodified GP sample demonstrated better stability and higher emission current than the boron nitride modified sample. Both samples deintercalated above 1,000 K.

Dependence of heavy metal burnup on nuclear data libraries for fast reactors

CERN Document Server

Ohki, S

2003-01-01

Japan Nuclear Cycle Development Institute (JNC) is considering the highly burnt fuel as well as the recycling of minor actinide (MA) in the development of commercialized fast reactor cycle systems. Higher accuracy in burnup calculation is going to be required for higher mass plutonium isotopes ( sup 2 sup 4 sup 0 Pu, etc.) and MA nuclides. In the framework of research and development aiming at the validation and necessary improvements of fast reactor burnup calculation, we investigated the differences among the burnup calculation results with the major nuclear data libraries: JEF-2.2, ENDF/B-VI Release 5, JENDL-3.2, and JENDL-3.3. We focused on the heavy metal nuclides such as plutonium and MA in the central core region of a conventional sodium-cooled fast reactor. For main heavy metal nuclides ( sup 2 sup 3 sup 5 U, sup 2 sup 3 sup 8 U, sup 2 sup 3 sup 9 Pu, sup 2 sup 4 sup 0 Pu, and sup 2 sup 4 sup 1 Pu), number densities after 1-cycle burnup did not change over one or two percent. Library dependence was re...

Nitriding behavior of Ni and Ni-based binary alloys

Energy Technology Data Exchange (ETDEWEB)

Fonovic, Matej

2015-01-15

Gaseous nitriding is a prominent thermochemical surface treatment process which can improve various properties of metallic materials such as mechanical, tribological and/or corrosion properties. This process is predominantly performed by applying NH{sub 3}+H{sub 2} containing gas atmospheres serving as the nitrogen donating medium at temperatures between 673 K and 873 K (400 C and 600 C). NH{sub 3} decomposes at the surface of the metallic specimen and nitrogen diffuses into the surface adjacent region of the specimen whereas hydrogen remains in the gas atmosphere. One of the most important parameters characterizing a gaseous nitriding process is the so-called nitriding potential (r{sub N}) which determines the chemical potential of nitrogen provided by the gas phase. The nitriding potential is defined as r{sub N} = p{sub NH{sub 3}}/p{sub H{sub 2}{sup 3/2}} where p{sub NH{sub 3}} and p{sub H{sub 2}} are the partial pressures of the NH{sub 3} and H{sub 2} in the nitriding atmosphere. In contrast with nitriding of α-Fe where the nitriding potential is usually in the range between 0.01 and 1 atm{sup -1/2}, nitriding of Ni and Ni-based alloys requires employing nitriding potentials higher than 100 atm{sup -1/2} and even up to ∞ (nitriding in pure NH{sub 3} atmosphere). This behavior is compatible with decreased thermodynamic stability of the 3d-metal nitrides with increasing atomic number. Depending on the nitriding conditions (temperature, nitriding potential and treatment time), different phases are formed at the surface of the Ni-based alloys. By applying very high nitriding potential, formation of hexagonal Ni{sub 3}N at the surface of the specimen (known as external nitriding) leads to the development of a compound layer, which may improve tribological properties. Underneath the Ni{sub 3}N compound layer, two possibilities exist: (i) alloying element precipitation within the nitrided zone (known as internal nitriding) and/or (ii) development of metastable and

Supramolecular intermediates in the synthesis of polymeric carbon nitride from melamine cyanurate

International Nuclear Information System (INIS)

Dante, Roberto C.; Sánchez-Arévalo, Francisco M.; Chamorro-Posada, Pedro; Vázquez-Cabo, José; Huerta, Lazaro; Lartundo-Rojas, Luis; Santoyo-Salazar, Jaime

2015-01-01

The adduct of melamine and cyanuric acid (MCA) was used in past research to produce polymeric carbon nitride and precursors. The reaction yield was considerably incremented by the addition of sulfuric acid. The polymeric carbon nitride formation occurs around 450 °C at temperatures above the sublimation of the adduct components, which occurs around 400 °C. In this report the effect of sulfuric acid on MCA was investigated. It was found that the MCA rosette supramolecular channel structures behave as a solid solvent able to host small molecules, such as sulfuric acid, inside these channels and interact with them. Therefore, the sulfuric acid effect was found to be close to that of a solute that causes a temperature increment of the “solvent sublimation” enough to allowing the formation of polymeric carbon nitride to occur. Sulfate ions are presumably hosted in the rosette channels of MCA as shown by simulations. - Graphical abstract: The blend of melamine cyanurate and sulfuric acid behaves like a solution so that melamine cyanurate decomposition is shifted to temperatures high enough to react and form polymeric carbon nitride. - Highlights: • The adduct of melamine and cyanuric acid behaves as a solid solvent. • The blend of sulfuric acid and melamine cyanurate behaves like a solution. • Melamine cyanurate decomposition is shifted to higher temperatures by sulfuric acid. • The formation of polymeric carbon nitride occurs for these higher temperatures

Supramolecular intermediates in the synthesis of polymeric carbon nitride from melamine cyanurate

Energy Technology Data Exchange (ETDEWEB)

Dante, Roberto C., E-mail: rcdante@yahoo.com [Facultad de Mecánica, Escuela Politécnica Nacional (EPN), Ladrón de Guevara E11-253, Quito (Ecuador); Sánchez-Arévalo, Francisco M. [Instituto de Investigaciones en Materiales, Universidad Nacional Autónoma de Mexico, Apdo. Postal 70-360, Cd. Universitaria, Mexico D.F. 04510 (Mexico); Chamorro-Posada, Pedro [Dpto. de Teoría de la Señal y Comunicaciones e IT, Universidad de Valladolid, ETSI Telecomunicación, Paseo Belén 15, 47011 Valladolid (Spain); Vázquez-Cabo, José [Dpto. de Teoría de la Señal y Comunicaciones, Universidad de Vigo, ETSI Telecomunicación, Lagoas Marcosende s/n, Vigo (Spain); Huerta, Lazaro [Instituto de Investigaciones en Materiales, Universidad Nacional Autónoma de Mexico, Apdo. Postal 70-360, Cd. Universitaria, Mexico D.F. 04510 (Mexico); Lartundo-Rojas, Luis [Centro de Nanociencias y Micro y Nanotecnologías—IPN, Luis Enrique Erro s/n, U. Prof. Adolfo López Mateos, 07738 Ciudad de Mexico, Distrito Federal (Mexico); Santoyo-Salazar, Jaime [Departamento de Física, Centro de Investigación y de Estudios Avanzados del Instituto Politécnico Nacional, CINVESTAV-IPN, Apdo. Postal 14-740, Mexico D.F. 07360 (Mexico); and others

2015-03-15

The adduct of melamine and cyanuric acid (MCA) was used in past research to produce polymeric carbon nitride and precursors. The reaction yield was considerably incremented by the addition of sulfuric acid. The polymeric carbon nitride formation occurs around 450 °C at temperatures above the sublimation of the adduct components, which occurs around 400 °C. In this report the effect of sulfuric acid on MCA was investigated. It was found that the MCA rosette supramolecular channel structures behave as a solid solvent able to host small molecules, such as sulfuric acid, inside these channels and interact with them. Therefore, the sulfuric acid effect was found to be close to that of a solute that causes a temperature increment of the “solvent sublimation” enough to allowing the formation of polymeric carbon nitride to occur. Sulfate ions are presumably hosted in the rosette channels of MCA as shown by simulations. - Graphical abstract: The blend of melamine cyanurate and sulfuric acid behaves like a solution so that melamine cyanurate decomposition is shifted to temperatures high enough to react and form polymeric carbon nitride. - Highlights: • The adduct of melamine and cyanuric acid behaves as a solid solvent. • The blend of sulfuric acid and melamine cyanurate behaves like a solution. • Melamine cyanurate decomposition is shifted to higher temperatures by sulfuric acid. • The formation of polymeric carbon nitride occurs for these higher temperatures.

Status of burnup credit for transport of SNF in the United States

International Nuclear Information System (INIS)

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

2004-01-01

Allowing credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transportation, and disposal of spent nuclear fuel (SNF) while maintaining a subcritical margin sufficient to establish an adequate safety basis. This paper reviews the current status of burnup credit applied to the design and transport of SNF casks in the United States. The existing U.S. regulatory guidance on burnup credit is limited to pressurized-water-reactor (PWR) fuel and to allowing credit only for actinides in the SNF. By comparing loading curves against actual SNF discharge data for U.S. reactors, the potential benefits that can be realized using the current regulatory guidance with actinide-only burnup credit are illustrated in terms of the inventory allowed in high-capacity casks and the concurrent reduction in SNF shipments. The additional benefits that might be realized by extending burnup credit to credit for select fission products are also illustrated. The curves show that, although fission products in SNF provide a small decrease in reactivity compared with actinides, the additional negative reactivity causes the SNF inventory acceptable for transportation to increase from roughly 30% to approximately 90% when fission products are considered. A savings of approximately $150M in transport costs can potentially be realized for the planned inventory of the repository. Given appropriate experimental data to support code validation, a realistic best-estimate analysis of burnup credit that includes validated credit for fission products is the enhancement that will yield the most significant impact on future transportation plans

Benefits of the delta K of depletion benchmarks for burnup credit validation

International Nuclear Information System (INIS)

Lancaster, D.; Machiels, A.

2012-01-01

Pressurized Water Reactor (PWR) burnup credit validation is demonstrated using the benchmarks for quantifying fuel reactivity decrements, published as 'Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty,' EPRI Report 1022909 (August 2011). This demonstration uses the depletion module TRITON available in the SCALE 6.1 code system followed by criticality calculations using KENO-Va. The difference between the predicted depletion reactivity and the benchmark's depletion reactivity is a bias for the criticality calculations. The uncertainty in the benchmarks is the depletion reactivity uncertainty. This depletion bias and uncertainty is used with the bias and uncertainty from fresh UO 2 critical experiments to determine the criticality safety limits on the neutron multiplication factor, k eff . The analysis shows that SCALE 6.1 with the ENDF/B-VII 238-group cross section library supports the use of a depletion bias of only 0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertainty in the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the delta k of depletion ('Kopp memo') was shown to be conservative for fuel with more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertainty is not a function of burnup, the Kopp memo's recommended bias and uncertainty may be exceeded at low burnups, but its absolute magnitude is small. (authors)

Design and analytic evaluation of a rim effect reduction type LWR fuel for extending burnup

International Nuclear Information System (INIS)

Matsumura, Tetsuo; Kameyama, Takanori; Kinoshita, Motoyasu

1991-01-01

We have designed a new concept fuel design 'Rim effect reduction type fuel' which has thin natural UO 2 layer on surface of a UO2 pellet. Our neutronic analyses with ANRB code show this fuel design can reduce rim effect (burnup at plelet rim) by about 30 GWd/t comparing a normal fuel. It is known that a high burnup fuel has different microstructure from as-fabricated one at fuel rim (which is called as rim region) due to rim effect. Therefore this fuel design can expect smaller rim region than a normal fuel. Our fuel performance analyses with EIMUS code show this fuel design can reduce fuel center temperature at high burnup if thermal conductivity of fuel pellet decreases with burnup in inverse proportion. However, this fuel design increases fuel center temperature at low and middle burnup than a normal fuel due to increase of thermal power density at pellet center. Additionally Irradiation experiment of this fuel design can be considered to offer important data which make clear the relation between rim effect and fuel performance. (author)

Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

International Nuclear Information System (INIS)

Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

2016-01-01

Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

Broadband wavelength conversion in hydrogenated amorphous silicon waveguide with silicon nitride layer

Science.gov (United States)

Wang, Jiang; Li, Yongfang; Wang, Zhaolu; Han, Jing; Huang, Nan; Liu, Hongjun

2018-01-01

Broadband wavelength conversion based on degenerate four-wave mixing is theoretically investigated in a hydrogenated amorphous silicon (a-Si:H) waveguide with silicon nitride inter-cladding layer (a-Si:HN). We have found that enhancement of the non-linear effect of a-Si:H waveguide nitride intermediate layer facilitates broadband wavelength conversion. Conversion bandwidth of 490 nm and conversion efficiency of 11.4 dB were achieved in a numerical simulation of a 4 mm-long a-Si:HN waveguide under 1.55 μm continuous wave pumping. This broadband continuous-wave wavelength converter has potential applications in photonic networks, a type of readily manufactured low-cost highly integrated optical circuits.

Technical and economic limits to fuel burnup extension. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2002-07-01

For many years, the increase of efficiency in the production of nuclear electricity has been an economic challenge in many countries which have developed this kind of energy. The increase of fuel burnup leads to a reduction in the volume of spent fuel discharged to longer fuel cycles in the reactor, which means bigger availability and capacity factors. After having increased the authorized burnup in plants, developing new alloys capable of resisting high burnup, and having accumulated data on fuel evolution with burnup, it has become necessary to establish the limitations which could be imposed by the physical evolution of the fuel, influencing fuel management, neutron properties, reprocessing or, more generally, the management of waste and irradiated fuels. It is also necessary to verify whether the benefits of lower electricity costs would not be offset by an increase in fuel management costs. The main questions are: Are technical and economic limits to the increasing of fuel burnup in parallel? Can we envisage nowadays the hardest limitation in some of these areas? Which are the main points to be solved from the technical point of view? Is this effort worthwhile considering the economy of the cycle? To which extent? For these reasons, the IAEA, following a recommendation by the International Working Group on Fuel Performance and Technology, held a Technical Committee Meeting on Technical and Economic Limits to Fuel Burnup Extension. The purpose of this meeting was to provide an international forum to review the evolution of fuel properties at increased burnup in order to estimate the limitations both from a physical and an economic point of view. The meeting was therefore divided into two parts. The first part, focusing on technical limits, was devoted to the improvement of the fuel element, such as fission gas release (FGR), RIM effect, cladding, etc. and the fabrication, core management, spent fuel and reprocessing. Eighteen related papers were presented which

Development of a Burnup Module DECBURN Based on the Krylov Subspace Method

Energy Technology Data Exchange (ETDEWEB)

Cho, J. Y.; Kim, K. S.; Shim, H. J.; Song, J. S

2008-05-15

This report is to develop a burnup module DECBURN that is essential for the reactor analysis and the assembly homogenization codes to trace the fuel composition change during the core burnup. The developed burnup module solves the burnup equation by the matrix exponential method based on the Krylov Subspace method. The final solution of the matrix exponential is obtained by the matrix scaling and squaring method. To develop DECBURN module, this report includes the followings as: (1) Krylov Subspace Method for Burnup Equation, (2) Manufacturing of the DECBURN module, (3) Library Structure Setup and Library Manufacturing, (4) Examination of the DECBURN module, (5) Implementation to the DeCART code and Verification. DECBURN library includes the decay constants, one-group cross section and the fission yields. Examination of the DECBURN module is performed by manufacturing a driver program, and the results of the DECBURN module is compared with those of the ORIGEN program. Also, the implemented DECBURN module to the DeCART code is applied to the LWR depletion benchmark and a OPR-1000 pin cell problem, and the solutions are compared with the HELIOS code to verify the computational soundness and accuracy. In this process, the criticality calculation method and the predictor-corrector scheme are introduced to the DeCART code for a function of the homogenization code. The examination by a driver program shows that the DECBURN module produces exactly the same solution with the ORIGEN program. DeCART code that equips the DECBURN module produces a compatible solution to the other codes for the LWR depletion benchmark. Also the multiplication factors of the DeCART code for the OPR-1000 pin cell problem agree to the HELIOS code within 100 pcm over the whole burnup steps. The multiplication factors with the criticality calculation are also compatible with the HELIOS code. These results mean that the developed DECBURN module works soundly and produces an accurate solution

Ion nitridation - physical and technological aspects

International Nuclear Information System (INIS)

Elbern, A.W.

1980-01-01

Ion nitridation, is a technique which allows the formation of a controlled thickness of nitrides in the surface of the material, using this material as the cathode in a low pressure glow discharge, which presents many advantages over the conventional method. A brief review of the ion nitriding technique, the physical fenomena involved, and we discuss technological aspects of this method, are presented. (Author) [pt

Silicon nitride-fabrication, forming and properties

International Nuclear Information System (INIS)

Yehezkel, O.

1983-01-01

This article, which is a literature survey of the recent years, includes description of several methods for the formation of silicone nitride, and five methods of forming: Reaction-bonded silicon nitride, sintering, hot pressing, hot isostatic pressing and chemical vapour deposition. Herein are also included data about mechanical and physical properties of silicon nitride and the relationship between the forming method and the properties. (author)

Fission product margin in burnup credit analyses

International Nuclear Information System (INIS)

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

1998-01-01

The US Department of Energy (DOE) is currently working toward the licensing of a methodology for using actinide-only burnup credit for the transportation of spent nuclear fuel (SNF). Important margins are built into this methodology. By using comparisons with a representative experimental database to determine bias factors, the methodology ensures that actinide concentrations and worths are estimated conservatively; furthermore, the negative net reactivity of certain actinides and all fission products (FPs) is not taken into account, thus providing additional margin. A future step of DOE's effort might aim at establishing an actinide and FP burnup credit methodology. The objective of this work is to establish the uncertainty to be applied to the total FP worth in SNF. This will serve two ends. First, it will support the current actinide-only methodology by demonstrating the margin available from FPs. Second, it will identify the major contributions to the uncertainty and help set priorities for future work

Effect of fissile isotope burnup on criticality safety for stored disintegrated fuel rods

International Nuclear Information System (INIS)

Heaberlin, S.W.; Selby, G.P.

1978-09-01

If the fuel rods were to disintegrate and water added, a criticality could occur in a 13-in. PWR canister with fresh fuel enriched to 3.5 wt % 235 U. The question is, ''If credit could be taken for burnup, could this indicate a subcritical condition.'' In attempting to answer this question, a series of calculations were performed. A set of isotopic concentrations were generated for 5,000, 10,000, 15,000, and 20,000 MWD/MTU burnup levels. Four reflector materials, water, concrete and two types of soil, were considered. Results indicate that allowing credit for fissile isotope burnup does not completely remove the concern for criticality safety in the event of rod disintegration. Reactivities which are ''subcritical'' (k/sub eff/ = 0.95) would not occur for three of the four reflector materials at even the 20,000 MWD/MTU burnup level in the 13-in. canister. The water reflected canister would achieve the k/sub eff/ = 0.95 level near 18,000 MWD/MTU. A smaller canister could be postulated. If a quarter inch gap is allowed, a Westinghouse 17 x 17 PWR assembly requires a 12 1 / 4 inch diameter canister. For such a canister with water reflection the ''subcritical'' (k/sub eff/ = 0.95) level would be reached near 15,000 MWD/MTU. The soil reflected canisters would reach this level between 18,000 and 19,000 MWD/MTU. Considering the difficulties in taking credit for burnup, such modest gains in apparent safety are not encouraging. This situation might be improved, however, if credit were also taken for neutron absorption by fission product poisons produced during burnup. It is strongly recommended that other approaches to a solution of the criticality safety problem be considered

Synthesis of reduced carbon nitride at the reduction by hydroquinone of water-soluble carbon nitride oxide (g-C{sub 3}N{sub 4})O

Energy Technology Data Exchange (ETDEWEB)

Kharlamov, Alexey [Frantsevich Institute for Problems of Materials Science of NASU, Krzhyzhanovsky St. 3, 03680 Kiev (Ukraine); Bondarenko, Marina, E-mail: mebondarenko@ukr.net [Frantsevich Institute for Problems of Materials Science of NASU, Krzhyzhanovsky St. 3, 03680 Kiev (Ukraine); Kharlamova, Ganna [Taras Shevchenko National University of Kiev, Volodymyrs' ka St. 64, 01601 Kiev (Ukraine); Fomenko, Veniamin [Frantsevich Institute for Problems of Materials Science of NASU, Krzhyzhanovsky St. 3, 03680 Kiev (Ukraine)

2016-09-15

For the first time at the reduction by hydroquinone of water-soluble carbon nitride oxide (g-C{sub 3}N{sub 4})O reduced carbon nitride (or reduced multi-layer azagraphene) is obtained. It is differed from usually synthesized carbon nitride by a significantly large (on 0.09 nm) interplanar distance is. At the same time, the chemical bonds between atoms in a heteroatomic plane of reduced carbon nitride correspond to the bonds in a synthesized g-C{sub 3}N{sub 4}. The samples of water-soluble carbon nitride oxide were synthesized under the special reactionary conditions of a pyrolysis of melamine and urea. We believe that reduced carbon nitride consists of weakly connected carbon-nitrogen monosheets (azagraphene sheets) as well as reduced (from graphene oxide) graphene contains weakly connected graphene sheets. - Graphical abstract: XRD pattern and schematic atomic model of one layer of reduced carbon nitride, carbon nitride oxide and synthesized carbon nitride. For the first time at the reduction by hydroquinone of the water-soluble carbon nitride oxide (g-C{sub 3}N{sub 4})O is obtained the reduced carbon nitride (or reduced multi-layer azagraphene). Display Omitted - Highlights: • First the reduced carbon nitride (RCN) at the reduction of the carbon nitride oxide was obtained. • Water-soluble carbon nitride oxide was reduced by hydroquinone. • The chemical bonds in a heteroatomic plane of RCN correspond to the bonds in a synthesized g-C{sub 3}N{sub 4}. • Reduced carbon nitride consists of poorly connected heteroatomic azagraphene layers.

Method for adding additional isotopes to actinide-only burnup credit

International Nuclear Information System (INIS)

Lancaster, D.B.; Fuentes, E.; Kang, C.

1998-01-01

The Topical Report on Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel Packages requires computer code validation to be performed against a benchmark set of chemical assays for isotopic concentration and against a benchmark set of critical experiments for package criticality. Both sets contain all the isotopes included in the methodology. The chemical assays used include the uranium and plutonium isotopes, while the critical experiments were composed of UO 2 or MOX rods, covering the isotopes in the actinide only approach. Since other isotopes are not included in the validation benchmark sets, it would be necessary to justify both the content and worth of any additional isotope for which burnup credit is to be taken (i.e., both the concentration and criticality effect of each particular isotope must be validated). A method is proposed here that can be used for any number of additional isotopes. As does the actinide-only burnup credit methodology, this method makes use of chemical assay data to establish the conservatism in the prediction of each isotope's concentration. Criticality validation is also performed using a benchmark set of UO 2 and MOX critical experiments, where the additional isotopes are validated using worth experiments to conservatively account for any uncertainty in their cross sections. The remaining requirements (analysis and modeling parameters, loading criteria generation, and physical implementation and controls) are performed exactly as described in the actinide-only burnup credit methodology. This report provides insight into each particular requirement in the new methodology

Overview of the burnup credit activities at OECD/NEA/NSC

International Nuclear Information System (INIS)

Brady Raap, M.C.; Nomura, Y.; Sartori, E.

2001-01-01

This article summarizes activities of the OECD/NEA Burnup Credit Expert Panel, a subordinate group to the Working Party on Nuclear Criticality Safety (WPNCS). The WPNCS of the OECD/NEA coordinates and carries out work in the domain of criticality safety at the international level. Particular attention is devoted to establishing sound databases required in this area and to addressing issues of high relevance such as burnup credit. The activities of the expert panel are aimed toward improving safety and identifying economic solutions to issues concerning the back-end of the fuel cycle. The main objective of the activities of the OECD/NEA Burnup Credit Expert Panel is to demonstrate that the available criticality safety calculational tools are appropriate for application to burned fuel systems and that a reasonable safety margin can be established. The method established by the expert panel for investigating the physics and predictability of burnup credit is based on the specification and comparison of calculational benchmark problems. A wide range of fuel types, including PWR, BWR, MOX, and VVER fuels, has been or are being addressed by the expert panel. The objective and status of each of these benchmark problems is reviewed in this article. It is important to note that the focus of the expert panel is the comparison of the results submitted by each participant to assess the capability of commonly used code systems, not to quantify the physical phenomena investigated in the comparisons or to make recommendations for licensing action. (author)

Burnup verification measurements at a US nuclear utility using the FORK measurement system

International Nuclear Information System (INIS)

Ewing, R.I.; Bosler, G.E.; Walden, G.

1993-01-01

The FORK measurement system, designed at Los Alamos National Laboratory (LANL) for the International Atomic Energy Agency (IAEA) safeguards program, has been used to examine spent reactor fuel assemblies at Duke Power Company's Oconee Nuclear Station. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. These measurements can be correlated with burnup and cooling time, and can be used to verify the reactor site records. Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. By taking into account the reduced reactivity of spent fuel due to its burnup in the reactor, burnup credit results in more efficient and economic transport and storage. The objectives of these tests are to demonstrate the applicability of the FORK system to verify reactor records and to develop optimal procedures compatible with utility operations. The test program is a cooperative effort supported by Sandia National Laboratories, the Electric Power Research Institute (EPRI), Los Alamos National Laboratory, and the Duke Power Company

A SAS2H/KENO-V methodology for 3D fuel burnup analysis

International Nuclear Information System (INIS)

Milosevic, M.; Greenspan, E.; Vujic, J.

2002-01-01

An efficient methodology for 3D fuel burnup analysis of LWR reactors is described in this paper. This methodology is founded on coupling Monte Carlo method for 3D calculation of node power distribution, and transport method for depletion calculation in ID Wigner-Seitz equivalent cell for each node independently. The proposed fuel burnup modeling, based on application of SCALE-4.4a control modules SAS2H and KENO-V.a is verified for the case of 2D x-y model of IRIS 15 x 15 fuel assembly (with reflective boundary condition) by using two well benchmarked code systems. The one is MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility code, and the second is KENO-V.a/ORIGEN2.1 code system recently developed by authors of this paper. The proposed SAS2H/KENO-V.a methodology was applied for 3D burnup analysis of IRIS-1000 benchmark.44 core. Detailed k sub e sub f sub f and power density evolution with burnup are reported. (author)

Proceedings of a workshop on the use of burnup credit in spent fuel transport casks

International Nuclear Information System (INIS)

Sanders, T.L.

1989-10-01

The Department of Energy sponsored a workshop on the use of burnup credit in the criticality design of spent fuel shipping casks on February 21 and 22, 1988. Twenty-five different presentations on many related topics were conducted, including the effects of burnup credit on the design and operation of spent fuel storage pools, casks and modules, and shipping casks; analysis and physics issues related to burnup credit; regulatory issues and criticality safety; economic incentives and risks associated with burnup credit; and methods for verifying spent fuel characteristics. An abbreviated version of the DOE workshop was repeated as a special session at the November 1988 American Nuclear Society Meeting in Washington, DC. Each of the invited speakers prepared detailed papers on his or her respective topic. The individual papers have been cataloged separately

The application of burnup credit for spent fuel operations in the United Kingdom

International Nuclear Information System (INIS)

Bowden, R.

1998-01-01

This paper begins by outlining the structure of the nuclear industry in the United Kingdom. It then sets out the methodology of burnup credit, and provides a brief discussion of the validation and robustness of the calculational route. This leads to a description of both the current and intended applications of burnup credit in the United Kingdom. (author)

Topotactic synthesis of vanadium nitride solid foams

International Nuclear Information System (INIS)

Oyama, S.T.; Kapoor, R.; Oyama, H.T.; Hofmann, D.J.; Matijevic, E.

1993-01-01

Vanadium nitride has been synthesized with a surface area of 120 m 2 g -1 by temperature programmed nitridation of a foam-like vanadium oxide (35 m 2 g -1 ), precipitated from vanadate solutions. The nitridation reaction was established to be topotactic and pseudomorphous by x-ray powder diffraction and scanning electron microscopy. The crystallographic relationship between the nitride and oxide was {200}//{001}. The effect of precursor geometry on the product size and shape was investigated by employing vanadium oxide solids of different morphologies

Performance of Bruce natural UO2 fuel irradiated to extended burnups

International Nuclear Information System (INIS)

Zhou, Y.N.; Floyd, M.R.; Ryz, M.A.

1995-11-01

Bruce-type bundles XY, AAH and GF were successfully irradiated in the NRU reactor at Chalk River Laboratories to outer-element burnups of 570-900 MWh/kgU. These bundles were of the Bruce Nuclear Generating Station (NGS)-A 'first-charge' design that contained gas plenums in the outer elements. The maximum outer-element linear powers were 33-37 kW/m. Post-irradiation examination of these bundles confirmed that all the elements were intact. Bundles XY and AAH, irradiated to outer-element burnups of 570-700 MWh/kgU, experienced low fission-gas release (FGR) ( 500 MWh/kgU (equivalent to bundle-average 450 MWh/kgU) when maximum outer-element linear powers are > 50 kW/m. The analysis in this paper suggests that CANDU 37-element fuel can be successfully irradiated (low-FGR/defect-free) to burnups of at least 700 MWh/kgU, provided maximum power do not exceed 40 kW/m. (author). 5 refs., 1 tab., 8 figs

Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

Energy Technology Data Exchange (ETDEWEB)

Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

2015-01-01

Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k_eff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

Microstructure and mechanical properties of silicon nitride structural ceramics of silicon nitride

International Nuclear Information System (INIS)

Strohaecker, T.R.; Nobrega, M.C.S.

1989-01-01

The utilization of direct evaluation technic of tenacity for fracturing by hardness impact in silicon nitride ceramics is described. The microstructure were analysied, by Scanning Electron Microscopy, equiped with a microanalysis acessory by X ray energy dispersion. The difference between the values of K IC measure for two silicon nitride ceramics is discussed, in function of the microstructures and the fracture surfaces of the samples studied. (C.G.C.) [pt

Anti corrosion layer for stainless steel in molten carbonate fuel cell - comprises phase vapour deposition of titanium nitride, aluminium nitride or chromium nitride layer then oxidising layer in molten carbonate electrolyte

DEFF Research Database (Denmark)

2000-01-01

Forming an anticorrosion protective layer on a stainless steel surface used in a molten carbonate fuel cell (MCFC) - comprises the phase vapour deposition (PVD) of a layer comprising at least one of titanium nitride, aluminium nitride or chromium nitride and then forming a protective layer in situ...