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Sample records for detailed burnup calculations

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)
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
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
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
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)
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
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)
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
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)
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)
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
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)
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)
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
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.
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)
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)
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)
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 calculation of a CANDU6 reactor using the Serpent and MCNP6 codes

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S.; Bonin, H.W., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, ON (Canada)

2014-07-01

A study of fuel burnup for the CANDU6 reactor is carried out to validate the most recent versions of the probabilistic transport code (MCNP6) and the continuous energy burnup calculation code (Serpent). These two codes allow for 3-D geometry calculation accounting for a detailed analysis without unit-cell homogenization. On the other hand, the WIMS-AECL computer program is used to model neutron transport in nuclear-reactor lattices for design, safety analysis, and operation. It works with two-dimensional regions and can perform collision probability calculations for a periodic structure of the lattice cell. In the present work, the multiplication factor, the total flux and fuel burnup could be calculated for a CANDU6 nuclear reactor based on the GENTILLY-2 core design. The MCNP6 and Serpent codes provide a calculation of the track length estimated flux per neutron source. This estimated flux is then scaled with normalization to the reactor power in order to provide a flux in unit of n/cm{sup 2}s. Good agreement is observed between the actual total flux calculated by MCNP6, Serpent and WIMS-AECL. The effective multiplication factors of the whole core CANDU6 reactor are further calculated as a function of burnup and further compared to those calculated by WIMS-AECL where excellent agreement is also obtained. (author)
Burnup calculation of a CANDU6 reactor using the Serpent and MCNP6 codes

International Nuclear Information System (INIS)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

2014-01-01

A study of fuel burnup for the CANDU6 reactor is carried out to validate the most recent versions of the probabilistic transport code (MCNP6) and the continuous energy burnup calculation code (Serpent). These two codes allow for 3-D geometry calculation accounting for a detailed analysis without unit-cell homogenization. On the other hand, the WIMS-AECL computer program is used to model neutron transport in nuclear-reactor lattices for design, safety analysis, and operation. It works with two-dimensional regions and can perform collision probability calculations for a periodic structure of the lattice cell. In the present work, the multiplication factor, the total flux and fuel burnup could be calculated for a CANDU6 nuclear reactor based on the GENTILLY-2 core design. The MCNP6 and Serpent codes provide a calculation of the track length estimated flux per neutron source. This estimated flux is then scaled with normalization to the reactor power in order to provide a flux in unit of n/cm 2 s. Good agreement is observed between the actual total flux calculated by MCNP6, Serpent and WIMS-AECL. The effective multiplication factors of the whole core CANDU6 reactor are further calculated as a function of burnup and further compared to those calculated by WIMS-AECL where excellent agreement is also obtained. (author)
Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

International Nuclear Information System (INIS)

Garcia-Herranz, Nuria; Cabellos, Oscar; Sanz, Javier; Juan, Jesus; Kuijper, Jim C.

2008-01-01

Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files
Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

Energy Technology Data Exchange (ETDEWEB)

Garcia-Herranz, Nuria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain)], E-mail: nuria@din.upm.es; Cabellos, Oscar [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain); Sanz, Javier [Departamento de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, UNED (Spain); Juan, Jesus [Laboratorio de Estadistica, Universidad Politecnica de Madrid, UPM (Spain); Kuijper, Jim C. [NRG - Fuels, Actinides and Isotopes Group, Petten (Netherlands)

2008-04-15

Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.
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)
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
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)
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)
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
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 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)
Effect of high burn-up and MOX fuel on reprocessing, vitrification and disposal of PWR and BWR spent fuels based on accurate burn-up calculation

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-01

To examine the procedures of the reprocessing, the vitrification and the geologic disposal, precise burn-up calculation for high burn-up and MOX fuels has been performed for not only PWR but also BWR by using SWAT and SWAT2 codes which are the integrated bum-up calculation code systems combined with the bum-up calculation code, ORIGEN2, and the transport calculation code, SRAC (the collision probability method) or MVP (the continuous energy Monte Carlo method), respectively. The calculation results shows that all of the evaluated items (heat generation and concentrations of Mo and Pt) largely increase and those significantly effect to the current procedures of the vitrification and the geologic disposal. The calculation result by SWAT2 confirms that the bundle calculation is required for BWR to be discussed about those effects in details, especially for the MOX fuel. (authors)
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)
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
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)
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.
SWAT4.0 - The integrated burnup code system driving continuous energy Monte Carlo codes MVP, MCNP and deterministic calculation code SRAC

International Nuclear Information System (INIS)

Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki

2015-03-01

There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0. (author)
New Burnup Calculation System for Fusion-Fission Hybrid System

International Nuclear Information System (INIS)

Isao Murata; Shoichi Shido; Masayuki Matsunaka; Keitaro Kondo; Hiroyuki Miyamaru

2006-01-01

model and burnup calculation by ORIGEN2. The series calculation can be easily executed in one batch processing by a simple script command. The details of the calculation system for analysis of fusion-fission hybrid reactor especially concentrating on making of collapsed cross section for ORIGEN is to be presented in the conference together with some burnup calculation results of Japanese possible fusion fission hybrid reactor. (author)
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
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

The applicability of detailed process for neutron resonance absorption to neutronics analyses in LWR next generation fuels to extend burnup

International Nuclear Information System (INIS)

Kameyama, Takanori; Nauchi, Yasushi

2004-01-01

Neutronics analyses with detail processing for neutron resonance absorption in LWR next generation UOX and MOX fuels to extend burnup were performed based on the neutronic transport and burnup calculation. In the detailed processing, ultra-fine energy nuclear library and collision probabilities between neutron and U, Pu nuclides (actinide nuclides) are utilized for two-dimension geometry. In the usual simple processing (narrow resonance approximation), shielding factors and compensation equations for neutron resonance absorption are utilized. The results with detailed and simple processing were compared to clarify where the detailed processing is needed. The two processing caused difference of neutron multiplication factor by 0.5% at the beginning of irradiation, while the difference became smaller as burnup increased and was not significant at high burnup. The nuclide compositions of the fuel rods for main actinide nuclides were little different besides Cm isotopes by the processing, since the neutron absorption rate of 244 Cm became different. The detail processing is needed to evaluate the neutron emission rate in spent fuels. In the fuel assemblies, the distributions of rod power rates were not different within 0.5%, and the peak rates of fuel rod were almost the same by the two processing at the beginning of irradiation when the peak rate is the largest during the irradiation. The simple processing is also satisfied for safety evaluation based on the peak rate of rod power. The difference of local power densities in fuel pellets became larger as burnup increased, since the neutron absorption rate of 238 U in the peripheral region of pellets were significantly different by the two processing. The detail processing is needed to evaluate the fuel behavior at high burnup. (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)
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)
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)
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)
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
Burnup calculations for cadmium. A case study for HFR experiments

Energy Technology Data Exchange (ETDEWEB)

Pijlgroms, B.J.; Sciolla, C.M

2000-09-11

This report describes the pre-design burnup calculations performed for a cadmium shielded high fluence irradiation experiment in the HFR. The very high absorption cross section in cadmium causes problems in the calculations for two different reasons. Firstly, because of the large reaction rates the assumption that the flux and the cross sections remain piecewise constant is no longer true. Therefore the correct solution can only be obtained when using extremely small time steps which leads to excessive computing times. Secondly, the self-shielding in the cadmium becomes complete (black absorber) causing the depletion to progress in a shell-wise manner. As a consequence the depletion evolves nearly linear instead of exponential with time. Because of this the depletion codes are used in a regime for which these have not been designed leading to a systematic error. The analysis shows however that a good estimate for the burnup time can be obtained by extrapolation from calculations with practically sized time steps and a correction is derived to compensate the systematic error. The calculations were done using the OCTOPUS burnup code system, including the 3-D Monte-Carlo spectrum code MCNP-4B and the depletion code FISPACT-4.2. Verifications were performed with the WIMS code system. The first part of the report describes the study of the cadmium burnup calculations for a shielded steel sample with the emphasis on analyzing the requirements for obtaining the correct solution. The second part describes the time-dependent power production calculations with the steel replaced by lithium containing ceramic material such as to be used in the 'High Fluence Irradiation of Ceramics for Fusion' (HICU) experiment. 12 refs.
On the influence of spatial discretization in LWR steady state and burnup calculations with HELIOS 1.9

International Nuclear Information System (INIS)

Merk, B.; Weiss, F. P.

2009-01-01

Cell and burnup calculations are fundamental to all deterministic static and transient 3D full core calculations for different operational states of the reactor. The spatial discretization used for the cell and burnup calculations influences significantly the results of full integral transport solutions. The influence of the discretization on k inf is shown for the steady state case and the influence on the neutron spectrum is analyzed. Moreover, the differences in k inf are presented for different spatial discretization strategies in the burnup calculation of Uranium Oxide (UOX) fuel. The resulting different flux distributions cause significant changes in the isotopic densities. The influence of the discretization strategies on the calculation of homogenized few group cross-sections is investigated. This detailed discretization study demonstrates the need for sufficiently fine discretization to produce reliable and accurate results when using integral transport methods. In contrast to the currently used discretization schemes, refined discretization is especially important in the moderator region of the unit cell to reproduce the influence on the thermal neutron spectrum. Additionally, the need for sufficient discretization affects the idea of full core calculations based on integral transport methods since it has to be discussed whether it is worth to do full core calculations with reduced discretization when facing this strong discretization effect. The computer resources required for full core calculations with fine discretization are currently not available. (authors)
Evaluation of RSG-GAS Core Management Based on Burnup Calculation

International Nuclear Information System (INIS)

Lily Suparlina; Jati Susilo

2009-01-01

Evaluation of RSG-GAS Core Management Based on Burnup Calculation. Presently, U 3 Si 2 -Al dispersion fuel is used in RSG-GAS core and had passed the 60 th core. At the beginning of each cycle the 5/1 fuel reshuffling pattern is used. Since 52 nd core, operators did not use the core fuel management computer code provided by vendor for this activity. They use the manually calculation using excel software as the solving. To know the accuracy of the calculation, core calculation was carried out using two kinds of 2 dimension diffusion codes Batan-2DIFF and SRAC. The beginning of cycle burn-up fraction data were calculated start from 51 st to 60 th using Batan-EQUIL and SRAC COREBN. The analysis results showed that there is a disparity in reactivity values of the two calculation method. The 60 th core critical position resulted from Batan-2DIFF calculation provide the reduction of positive reactivity 1.84 % Δk/k, while the manually calculation results give the increase of positive reactivity 2.19 % Δk/k. The minimum shutdown margin for stuck rod condition for manual and Batan-3DIFF calculation are -3.35 % Δk/k dan -1.13 % Δk/k respectively, it means that both values met the safety criteria, i.e <-0.5 % Δk/k. Excel program can be used for burn-up calculation, but it is needed to provide core management code to reach higher accuracy. (author)
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)
Study of the acceleration of nuclide burnup calculation using GPU with CUDA

International Nuclear Information System (INIS)

Okui, S.; Ohoka, Y.; Tatsumi, M.

2009-01-01

The computation costs of neutronics calculation code become higher as physics models and methods are complicated. The degree of them in neutronics calculation tends to be limited due to available computing power. In order to open a door to the new world, use of GPU for general purpose computing, called GPGPU, has been studied [1]. GPU has multi-threads computing mechanism enabled with multi-processors which realize mush higher performance than CPUs. NVIDIA recently released the CUDA language for general purpose computation which is a C-like programming language. It is relatively easy to learn compared to the conventional ones used for GPGPU, such as OpenGL or CG. Therefore application of GPU to the numerical calculation became much easier. In this paper, we tried to accelerate nuclide burnup calculation, which is important to predict nuclides time dependence in the core, using GPU with CUDA. We chose the 4.-order Runge-Kutta method to solve the nuclide burnup equation. The nuclide burnup calculation and the 4.-order Runge-Kutta method were suitable to the first step of introduction CUDA into numerical calculation because these consist of simple operations of matrices and vectors of single precision where actual codes were written in the C++ language. Our experimental results showed that nuclide burnup calculations with GPU have possibility of speedup by factor of 100 compared to that with CPU. (authors)
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
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)
A new approach to make collapsed cross section for burnup calculation of subcritical system

International Nuclear Information System (INIS)

Matsunaka, Masayuki; Kondo, Keitaro; Miyamaru, Hiroyuki; Murata, Isao

2008-01-01

A general-purpose transport and burnup code system for precise analysis of subcritical reactors like a fusion-fission (FF) hybrid reactor was developed and used for analyzing their performance. The FF hybrid reactor is a subcritical system, which has a concept of fusion reactor with a blanket region containing nuclear fuel and has been under discussion by author's group for years because the present burnup calculation system mainly consists of a general-purpose Monte Carlo code MCNP-4B, a point burnup code ORIGEN2. JENDL-3.3 pointwise cross section library and JENDL Activation Cross Section File 96 were used as base cross section libraries to make group constant for burnup calculation. A new method has been proposed to make group constant for the burnup calculation as accurate as possible directly using output data of the neutron transport calculation by MCNP and evaluated nuclear data libraries. This method is strict and a general procedure to make one group cross sections in Monte Carlo calculations, while it takes very long computation time. Some speed-up techniques were discussed for the present group constant making process so as to decrease calculation time. Adoption of postprocessing to make group constant improved the calculation accuracy because of increasing number of cross sections to be updated in each burnup cycle. The present calculation system is capable of performing neutronics analysis of subcritical reactors more precise than our previous one. However, at the moment, it still takes long computation time to make group constants. Further speed-up techniques are now under investigation so as to apply the present system to neutronics design analysis for various subcritical systems. (author)
Determination of the burn-up of TRIGA fuel elements by calculation with new TRIGLAV program

International Nuclear Information System (INIS)

Zagar, T.; Ravnik, M.

1996-01-01

The results of fuel element burn-up calculations with new TRIGLAV program are presented. TRIGLAV program uses two dimensional model. Results of calculation are compared to results calculated with program, which uses one dimensional model. The results of fuel element burn-up measurements with reactivity method are presented and compared with the calculated results. (author)
Application of dynamic pseudo fission products and actinides for accurate burnup calculations

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Kloosterman, J.L.

1996-09-01

The introduction of pseudo fission products for accurate fine-group spectrum calculations during burnup is discussed. The calculation of the density of the pseudo nuclides is done before each spectrum calculation from the actual densities and their cross sections of all nuclides to be lumped into a pseudo fission product. As there are also many actinides formed in the fuel during its life cycle, a pseudo actinide with fission cross section is also introduced. From a realistic burnup calculation it is demonstrated that only a few fission products and actinides need to be included explicitly in a spectrum calculation. All other fission products and actinides can be accurately represented in the pseudo nuclides. (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)
Calculation of fuel burn-up and fuel reloading for the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Lan, Nguyen Phuoc; Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Binh, Do Quang [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

Calculation of fuel burnup and fuel reloading for the Dalat Nuclear Research Reactor was carried out by using a new programme named HEXA-BURNUP, realized in a PC. The programme is used to calculate the following parameters of the Dalat reactor: a/Critical configurations of the core loaded with 69, 72, 74, 86, 88, 89 and 92 fuel elements. The effective multiplication coefficients equal 1 within the error ranges of less than 0.38%. b/ The thermal neutron flux distribution in the reactor. The calculated results agree with the experimental data measured at 11 typical positions. c/The average fuel burn-up for the period from Feb. 1984 to Sep. 1992. The difference between calculation and experiment is only about 1.9%. 10 fuel reloading versions are calculated, from which an optimal version is proposed. (author). 9 refs., 4 figs., 5 tabs.
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
Estimation of the impact of manufacturing tolerances on burn-up calculations using Monte Carlo techniques

Energy Technology Data Exchange (ETDEWEB)

Bock, M.; Wagner, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH, Garching (Germany). Forschungszentrum

2012-11-01

In recent years, the availability of computing resources has increased enormously. There are two ways to take advantage of this increase in analyses in the field of the nuclear fuel cycle, such as burn-up calculations or criticality safety calculations. The first possible way is to improve the accuracy of the models that are analyzed. For burn-up calculations this means, that the goal to model and to calculate the burn-up of a full reactor core is getting more and more into reach. The second way to utilize the resources is to run state-of-the-art programs with simplified models several times, but with varied input parameters. This second way opens the applicability of the assessment of uncertainties and sensitivities based on the Monte Carlo method for fields of research that rely heavily on either high CPU usage or high memory consumption. In the context of the nuclear fuel cycle, applications that belong to these types of demanding analyses are again burn-up and criticality safety calculations. The assessment of uncertainties in burn-up analyses can complement traditional analysis techniques such as best estimate or bounding case analyses and can support the safety analysis in future design decisions, e.g. by analyzing the uncertainty of the decay heat power of the nuclear inventory stored in the spent fuel pool of a nuclear power plant. This contribution concentrates on the uncertainty analysis in burn-up calculations of PWR fuel assemblies. The uncertainties in the results arise from the variation of the input parameters. In this case, the focus is on the one hand on the variation of manufacturing tolerances that are present in the different production stages of the fuel assemblies. On the other hand, uncertainties that describe the conditions during the reactor operation are taken into account. They also affect the results of burn-up calculations. In order to perform uncertainty analyses in burn-up calculations, GRS has improved the capabilities of its general

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
Burnup calculation with estimated neutron spectrum of JMTR irradiation field. Development of the burnup calculation method for fuel pre-irradiated in the JMTR

International Nuclear Information System (INIS)

Okonogi, Kazunari; Nakamura, Takehiko; Yoshinaga, Makio; Hosoyamada, Ryuji

1999-03-01

As a series of the pulse irradiation tests with the irradiated fuel, the high-enriched fuel rods pre-irradiated in the JMTR as well as the fuels irradiated in commercial reactors have been irradiated in the NSRR. In the pre-irradiation at the JMTR, the test fuels were placed at the irradiation holes in the reflector region far from the driver core to keep the linear heat generation rate of the test fuel low. Accordingly, neutron energy spectra of the irradiation holes for the test fuels are softened due to the higher moderator ratio than in those of the ordinary LWR core, which causes quite different burnup characteristics. JMTR post irradiation condition corresponds to the pre-test condition in the NSRR. Therefore, proper understanding of the condition is quite important for the precise evaluating the energy deposition and FP generation in the test. Then, neutron spectra at the JMTR irradiation field were evaluated and its effects on the burnup calculation were quantified. Basing on the configuration of the JMTR core in the operation cycle No.85, neutron diffusion calculations of 107 groups were executed in 2-D slab (X-Y) geometry of CITATION of SRAC95 code system, and neutron energy spectra of the irradiation hole for the test fuels were evaluated. Burnup calculations of Test JMN-1 fuel with the estimated neutron energy spectra were performed and the results were compared to both the measurements and calculation results with the PWR and BWR libraries in ORIGEN2 code. SWAT code was used to collapse the 107 groups spectra into 1 group libraries for the ORIGEN2 use. The calculation results for both the generation and depletion of U, Pu and Nd with the JMTR libraries obtained in the present study were in the reasonably good agreement with the measurements, while in the case of calculation with the PWR and BWR libraries in ORIGEN2, the generation of fission products having mass numbers from 105 to 130 and some actinides were overestimated by about 1.5 to 3.5 times
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 neutronic calculations for research and training reactors via SCALE4.4

International Nuclear Information System (INIS)

Tombakoglu, M.; Cecen, Y.

2001-01-01

In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)
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.)
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.
MCNPX and MCB coupled methodology for the burnup calculation of the KIPT accelerator driven subcritical system

International Nuclear Information System (INIS)

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

2009-01-01

Argonne National Laboratory (ANL) of USA and Kharkov Inst. of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical facility (ADS). The facility will be utilized for basic research, medical isotopes production, and training young nuclear specialists. The burnup methodology and analysis of the KIPT ADS are presented in this paper. MCNPX and MCB Monte Carlo computer codes have been utilized. MCNPX has the capability of performing electron, photon and neutron coupled transport problems, but it lacks the burnup capability for driven subcritical systems. MCB has the capability for performing the burnup calculation of driven subcritical systems, while it cannot transport electrons. A calculational methodology coupling MCNPX and MCB has been developed, which can exploit the electrons transport capability of MCNPX for neutron production and the burnup capability of MCB for driven subcritical systems. In this procedure, a neutron source file is generated using MCNPX transport calculation, preserving the neutrons yield from photonuclear reactions initiated by electrons, and this source file is utilized by MCB for the burnup analyses with the same geometrical model. In this way, the ADS depletion calculation can be accurately. (authors)
Some implications of batch average burnup calculations on predicted spent fuel compositions

International Nuclear Information System (INIS)

Alexander, C.W.; Croff, A.G.

1984-01-01

The accuracy of using batch-averaged burnups to determine spent fuel characteristics (such as isotopic composition, activity, etc.) was examined for a typical pressurized-water reactor (PWR) fuel discharge batch by comparing characteristics computed by (a) performing a single depletion calculation using the average burnup of the spent fuel and (b) performing separate depletion calculations based on the relative amounts of spent fuel in each of twelve burnup ranges and summing the results. The computations were done using ORIGEN 2. Procedure (b) showed a significant shift toward a greater quantity of the heavier transuranics, which derive from multiple neutron captures, and a corresponding decrease in the amounts of lower transuranics. Those characteristics which derive primarily from fission products, such as total radioactivity and total thermal power, are essentially identical for the two procedures. Those characteristics that derive primarily from the heavier transuranics, such as spontaneous fission neutrons, are underestimated by procedure (a)
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)
Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

Energy Technology Data Exchange (ETDEWEB)

Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

2005-07-01

Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)
Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

International Nuclear Information System (INIS)

Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H.

2005-01-01

Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)
On the theories, techniques, and computer codes used in numerical reactor criticality and burnup calculations

International Nuclear Information System (INIS)

El-Osery, I.A.

1981-01-01

The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented
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)
Calculational prediction of fuel burn-up for the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Nguyen Phuoc Lan; Do Quang Binh

2016-01-01

In this paper, the method of expanding operators and functions in the neutron diffusion equations as chains of time variable is used for calculation of fuel burn-up of the Dalat nuclear reactors. A computer code, named BURREF, programmed in language Fortran-77 running on IBM PC-AT, has been developed based on this method to predict the fuel burn-up of the Dalat reactor. Some results will be presented here. (author)
Specification of phase 3 benchmark (Hex-Z heterogeneous and burnup calculation)

International Nuclear Information System (INIS)

Kim, Y.I.

2002-01-01

During the second RCM of the IAEA Co-ordinated Research Project Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects the following items were identified as important. Heterogeneity will affect absolute core reactivity. Rod worths could be considerably reduced by heterogeneity effects depending on their detailed design. Heterogeneity effects will affect the resonance self-shielding in the treatment of fuel Doppler, steel Doppler and sodium density effects. However, it was considered more important to concentrate on the sodium density effect in order to reduce the calculational effort required. It was also recognized that burnup effects will have an influence on fuel Doppler and sodium worths. A benchmark for the assessment of heterogeneity effect for Phase 3 was defined. It is to be performed for the Hex-Z model of the reactor only. No calculations will be performed for the R-Z model. For comparison with heterogeneous evaluations, the control rod worth will be calculated at the beginning of the equilibrium cycle, based on the homogeneous model. The definitions of rod raised and rod inserted for SHR are given, using the composition numbers
Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

Energy Technology Data Exchange (ETDEWEB)

Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

2014-05-01

The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.
Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

International Nuclear Information System (INIS)

Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh

2014-01-01

The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects
Details on an actinide-only burnup credit application in the USA

International Nuclear Information System (INIS)

Lancaster, D.

2001-01-01

Details on the Actinide-Only burnup credit assumptions that will be used for the CASTOR X/32 S cask are presented. Preliminary results show that using a conservative set of assumptions the cask will allow most fuel to be loaded without the addition of any additional reactivity control. With the addition of 8 control rod elements it is possible to load most of the rest of the fuel. (author)
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)
Re-evaluation of Assay Data of Spent Nuclear Fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya, E-mail: suyama.kenya@jaea.go.jp [Office of International Relations, Nuclear Safety Division, Ministry of Education, Culture, Sports, Science and Technology - Japan, 3-2-2 Kasumigaseki, Chiyoda-ku, Tokyo 100-8959 (Japan); Murazaki, Minoru; Ohkubo, Kiyoshi [Fuel Cycle Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency, 2-4 Shirakata Shirane, Tokai-mura, Ibaraki 319-1195 (Japan); Nakahara, Yoshinori [Research Group for Analytical Science, Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency, 2-4 Shirakata Shirane, Tokai-mura, Ibaraki 319-1195 (Japan); Uchiyama, Gunzo [Fuel Cycle Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency, 2-4 Shirakata Shirane, Tokai-mura, Ibaraki 319-1195 (Japan)

2011-05-15

Highlights: > The specifications required for the analyses of the destructive assay data taken from irradiated fuel in Ohi-1 and Ohi-2 PWRs were documented in this paper. > These data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. > These destructive assay data are suitable for the benchmarking of the burnup calculation code systems. - Abstract: The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries. In this paper, Assay Data of Spent Nuclear Fuel from two fuel assemblies irradiated in the Ohi-1 and Ohi-2 PWRs in Japan are shown. The destructive assay data from Ohi-2 have already been published. However, these data were not suitable for the benchmarking of calculation codes and libraries because several important specifications and data were not included. This paper summarizes the details of destructive assay data and specifications required for analyses of isotopic composition from Ohi-1 and Ohi-2. For precise burnup analyses, the burnup values of destructive assay samples were re-evaluated in this study. These destructive assay data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. This indicates that the quality of destructive assay data from Ohi-1 and Ohi-2 PWRs is high, and that these destructive assay data are suitable for the benchmarking of burnup calculation code systems.

Burn-up calculation of fusion-fission hybrid reactor using thorium cycle

International Nuclear Information System (INIS)

Shido, S.; Matsunaka, M.; Kondo, K.; Murata, I.; Yamamoto, Y.

2006-01-01

A burn-up calculation system has been developed to estimate performance of blanket in a fusion-fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In this system, neutron flux is calculated by MCNP4B and then burn-up calculation is performed by ORIGEN2. The cross-section library for ORIGEN2 is made from the calculated neutron flux and evaluated nuclear data. The 3-dimensional ITER model was used as a base fusion reactor. The nuclear fuel (reprocessed plutonium as the fission materials mixed with thorium as the fertile materials), transmutation materials (minor actinides and long-lived fission products) and tritium breeder were loaded into the blanket. Performances of gas-cooled and water-cooled blankets were compared with each other. As a result, the proposed reactor can meet the requirement for TBP and power density. As far as nuclear waste incineration is concerned, the gas-cooled blanket has advantages. On the other hand, the water cooled-blanket is suited to energy production. (author)
Advanced burnup calculation code system in a subcritical state with continuous-energy Monte Carlo code for fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Matsunaka, Masayuki; Ohta, Masayuki; Miyamaru, Hiroyuki; Murata, Isao

2009-01-01

The fusion-fission (FF) hybrid reactor is a promising energy source that is thought to act as a bridge between the existing fission reactor and the genuine fusion reactor in the future. The burnup calculation system that aims at precise burnup calculations of a subcritical system was developed for the detailed design of the FF hybrid reactor, and the system consists of MCNP, ORIGEN, and postprocess codes. In the present study, the calculation system was substantially modified to improve the calculation accuracy and at the same time the calculation speed as well. The reaction rate estimation can be carried out accurately with the present system that uses track-length (TL) data in the continuous-energy treatment. As for the speed-up of the reaction rate calculation, a new TL data bunching scheme was developed so that only necessary TL data are used as long as the accuracy of the point-wise nuclear data is conserved. With the present system, an example analysis result for our proposed FF hybrid reactor is described, showing that the computation time could really be saved with the same accuracy as before. (author)
OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

Energy Technology Data Exchange (ETDEWEB)

DeHart, M.D.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Labs., Las Vegas, NV (United States)

1996-06-01

In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, 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 agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155.
OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

International Nuclear Information System (INIS)

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

1996-06-01

In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, 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 agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155
Calculation of triton confinement and burn-up in tokamaks

International Nuclear Information System (INIS)

Anderson, D.; Battistoni, P.

1987-01-01

An analytical investigation is made of the confinement and subsequent burn-up of fusion produced tritons in a deuterium Tokamak plasma. Explicit approximations are obtained for the triton confinement factor, clearly displaying the scaling with physical parameters. The importance of pitch angle scattering losses during the triton slowing down is also estimated. A comparison with experiments and numerical calculations on the FT Tokamak slows good qualitative agreement. (authors)
FUEL BURN-UP CALCULATION FOR WORKING CORE OF THE RSG-GAS RESEARCH REACTOR AT BATAN SERPONG

Directory of Open Access Journals (Sweden)

Tukiran Surbakti

2017-12-01

Full Text Available The neutronic parameters are required in the safety analysis of the RSG-GAS research reactor. The RSG-GAS research reactor, MTR (Material Testing Reactor type is used for research and also in radioisotope production. RSG-GAS has been operating for 30 years without experiencing significant obstacles. It is managed under strict requirements, especially fuel management and fuel burn-up calculations. The reactor is operated under the supervision of the Regulatory Body (BAPETEN and the IAEA (International Atomic Energy Agency. In this paper, the experience of managing RSG-GAS core fuels will be discussed, there are hundred possibilities of fuel placements on the reactor core and the strategy used to operate the reactor will be crucial. However, based on strict calculation and supervision, there is no incorrect placement of the fuels in the core. The calculations were performed on working core by using the WIMSD-5B computer code with ENDFVII.0 data file to generate the macroscopic cross-section of fuel and BATAN-FUEL code were used to obtain the neutronic parameter value such as fuel burn-up fractions. The calculation of the neutronic core parameters of the RSG-GAS research reactor was carried out for U3Si2-Al fuel, 250 grams of mass, with an equilibrium core strategy. The calculations show that on the last three operating cores (T90, T91, T92, all fuels meet the safety criteria and the fuel burn-up does not exceed the maximum discharge burn-up of 59%. Maximum fuel burn-up always exists in the fuel which is close to the position of control rod.
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)
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
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)
Higher order methods for burnup calculations with Bateman solutions

International Nuclear Information System (INIS)

Isotalo, A.E.; Aarnio, P.A.

2011-01-01

Highlights: → Average microscopic reaction rates need to be estimated at each step. → Traditional predictor-corrector methods use zeroth and first order predictions. → Increasing predictor order greatly improves results. → Increasing corrector order does not improve results. - Abstract: A group of methods for burnup calculations solves the changes in material compositions by evaluating an explicit solution to the Bateman equations with constant microscopic reaction rates. This requires predicting representative averages for the one-group cross-sections and flux during each step, which is usually done using zeroth and first order predictions for their time development in a predictor-corrector calculation. In this paper we present the results of using linear, rather than constant, extrapolation on the predictor and quadratic, rather than linear, interpolation on the corrector. Both of these are done by using data from the previous step, and thus do not affect the stepwise running time. The methods were tested by implementing them into the reactor physics code Serpent and comparing the results from four test cases to accurate reference results obtained with very short steps. Linear extrapolation greatly improved results for thermal spectra and should be preferred over the constant one currently used in all Bateman solution based burnup calculations. The effects of using quadratic interpolation on the corrector were, on the other hand, predominantly negative, although not enough so to conclusively decide between the linear and quadratic variants.
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)
Isocrit: a burnup credit tool for spent fuel pool storage calculations - 333

International Nuclear Information System (INIS)

Kucukboyaci, V.N.; Marshall, W.J.

2010-01-01

In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse's state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power up-rate, exit temperature changes, etc) with a quick turnaround. (authors)
Confluence of calculational and experimental information for determination of power distribution and burnup

International Nuclear Information System (INIS)

Serov, I.V.; Hoogenboom, J.E.

1996-01-01

A technique for the statistical confluence of any number of possibly correlated informational sources employed in reactor analysis can be used to improve the estimates of physical quantities given by the sources taken separately. The formulas of the presented technique being based on multivariate Bayesian conditioning are general and can be employed in different applications. Insight into the nature of the informational source allows different types of data associated with the source to be improved. Estimation of biases, variances and correlation coefficients for the systematic and statistical errors associated with the informational sources is reliable confluence, but pays off by providing optimal estimates. The technique of the calculational and experimental information confluence is applied to the determination of the power distribution and burnup for the research reactor HOR of the Delft University of Technology. The code system CONHOR carries out all the stages of the calculation for the HOR reactor, using an existing code for static core calculations and burnup calculations. (author)
Confluence of calculational and experimental information for determination of power distribution and burnup

Energy Technology Data Exchange (ETDEWEB)

Serov, I.V.; Hoogenboom, J.E. [Interuniversitair Reactor Inst., Delft (Netherlands)

1996-05-01

A technique for the statistical confluence of any number of possibly correlated informational sources employed in reactor analysis can be used to improve the estimates of physical quantities given by the sources taken separately. The formulas of the presented technique being based on multivariate Bayesian conditioning are general and can be employed in different applications. Insight into the nature of the informational source allows different types of data associated with the source to be improved. Estimation of biases, variances and correlation coefficients for the systematic and statistical errors associated with the informational sources is reliable confluence, but pays off by providing optimal estimates. The technique of the calculational and experimental information confluence is applied to the determination of the power distribution and burnup for the research reactor HOR of the Delft University of Technology. The code system CONHOR carries out all the stages of the calculation for the HOR reactor, using an existing code for static core calculations and burnup calculations. (author).
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)
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)
A Study for Burn-up Calculation applied on 400MWth PBMR Core

International Nuclear Information System (INIS)

Luu, Nam Hai; Kim, Hong Chul; Kim, Soon Young; Kim, Jong Kyung; Noh, Jae Man

2007-01-01

The 400MWth Pebble-bed Modular Reactor (PBMR) is an advanced high temperature gas cooled-reactor (HTGR). It possesses a very high efficiency and attractive economics without compromising the high levels of passive safety expected of advanced nuclear designs. With this reason, PBMR is a target which researchers especially in nuclear engineering field study carefully and therefore it is regarded as the leader in the power generation field. There are many research results about benchmark problems but results of the burn-up process are still poor. Hence, in this study a burn-up calculation was performed with PBMR using MONTEBURNS code in which MCNP modeling linked a depletion systems is used
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)
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)
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
Calculation of burn-up data for spent LWR-fuels with respect to the design of spent fuel reprocessing plants

International Nuclear Information System (INIS)

Gasteiger, R.

1976-11-01

The design of spent fuel reprocessing plants makes necessary a detailed knowledge of the composition of the incoming fuels as a function of burn-up. This report gives a broad review on the composition of radionuclides in fuels (fission products, actinides) and structural materials for different burn-up data. (orig.) [de
Calculation Of Fuel Burnup And Radionuclide Inventory In The Syrian Miniature Neutron Source Reactor Using The GETERA Code

International Nuclear Information System (INIS)

Khattab, K.; Dawahra, S.

2011-01-01

Calculations of the fuel burnup and radionuclide inventory in the Syrian Miniature Neutron Source Reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer and has a bigger library of isotopes and more accurate. (author)
Nuclear-data uncertainty propagations in burnup calculation for the PWR assembly

International Nuclear Information System (INIS)

Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Shen, Wei

2017-01-01

Highlights: • The DRAGON 5.0 and NECP-CACTI have been implemented in UNICORN. • The effects of different neutronics methods on S&U results were quantified. • Uncertainty analysis has been applied to burnup calculation of PWR assembly. • The uncertainties of eigenvalue and few-group constants have been quantified. - Abstract: In this paper, our home-developed lattice code NECP-CACTI has been implemented into our UNICORN code to perform sensitivity and uncertainty analysis for the lattice calculations. The verified multigroup cross-section perturbation model and methods of the sensitivity and uncertainty analysis are established and applied to different lattice codes in UNICORN. As DRAGON5.0 and NECP-CACTI are available for the lattice calculations in UNICORN now, the effects of different neutronics methods (including methods for the neutron-transport and resonance self-shielding calculations) on the results of sensitivity and uncertainty analysis were studied in this paper. Based on NECP-CACTI, uncertainty analysis using the statistical sampling method has been performed to the burnup calculation for the fresh-fueled TMI-1 assembly, propagating the nuclear-data uncertainties to k_∞ and two-group constants of the lattice calculation with depletions. As results shown, for different neutronics methods, it can be observed that different methods of the neutron-transport calculation introduce no differences to the results of sensitivity and uncertainty analysis, while different methods of the resonance self-shielding calculation would impact the results. With depletions of the TMI-1 assembly, for k_∞, the relative uncertainty varies between 0.45% and 0.60%; for two-group constants, the largest variation is between 0.35% and 2.56% for vΣ_f_,_2. Moreover, the most significant contributors to the uncertainty of k_∞ and two-group constants varied with depletions are determined.
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
Burnable poisons in the light water reactor design, microburnup experiments and calculations. Part of a coordinated programme on burnup calculations and experiments for thermal reactors

International Nuclear Information System (INIS)

Penndorf, K.

1976-04-01

Investigations on Research Agreement N 1519/CF (1.8.1974 - 31.7.1975) entitled ''Burnable poisons in light water reactor design, microburnup experiments and calculations'' were carried out in the frame of the IAEA's coordinated research programme on ''Burn-up calculation and experiments for thermal reactors''. The theoretical and experimental work on application of solid burnable poison used for reduction of the amount of boric acid necessary to control of PWR or to lower the number of control rods needed in a BWR. Solid burnable poisons are needed in present PWR designs for the reduction of the boron acid concentration in order to prevent positive coefficients of reactivity. The special operational conditions of a ship reactor lead to the application of this kind of poison for compensation of almost all burnup reactivity. This strengthens the necessity of a very accurate and many dimensional calculations because an appropriate binding of reactivity has to be kept over the whole cycle time. Several burnup experiments had been run in the 15 MW material test reactor FRG-II. The following devices have been irradiated: poison pins within and without PWR fuel pin lattice segments and fuel pins containing pellets with a poison core. Measurements of reactivity, fluence, fission product concentration have been performed. Methods applied were γ-scanning and neutron pulse, radiography and transmission measurement techniques. Evaluation of the experiments was done by one and two dimensional Ssub(N) transport burnup calculations. In parallel a collision probability transport burnup code for current PWR design work is being developed, the main feature of which is economy in manpower and computer time
Validating analysis methodologies used in burnup credit criticality calculations

International Nuclear Information System (INIS)

Brady, M.C.; Napolitano, D.G.

1992-01-01

The concept of allowing reactivity credit for the depleted (or burned) state of pressurized water reactor fuel in the licensing of spent fuel facilities introduces a new challenge to members of the nuclear criticality community. The primary difference in this analysis approach is the technical ability to calculate spent fuel compositions (or inventories) and to predict their effect on the system multiplication factor. Isotopic prediction codes are used routinely for in-core physics calculations and the prediction of radiation source terms for both thermal and shielding analyses, but represent an innovation for criticality specialists. This paper discusses two methodologies currently being developed to specifically evaluate isotopic composition and reactivity for the burnup credit concept. A comprehensive approach to benchmarking and validating the methods is also presented. This approach involves the analysis of commercial reactor critical data, fuel storage critical experiments, chemical assay isotopic data, and numerical benchmark calculations
Development and verification of Monte Carlo burnup calculation system

International Nuclear Information System (INIS)

Ando, Yoshihira; Yoshioka, Kenichi; Mitsuhashi, Ishi; Sakurada, Koichi; Sakurai, Shungo

2003-01-01

Monte Carlo burnup calculation code system has been developed to evaluate accurate various quantities required in the backend field. From the Actinide Research in a Nuclear Element (ARIANE) program, by using, the measured nuclide compositions of fuel rods in the fuel assemblies irradiated in the commercial Netherlands BWR, the analyses have been performed for the code system verification. The code system developed in this paper has been verified through analysis for MOX and UO2 fuel rods. This system enables to reduce large margin assumed in the present criticality analysis for LWR spent fuels. (J.P.N.)
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)
Burnup credit calculations for criticality safety justification for RBMK-1000 spent fuel of transport and storage systems

Directory of Open Access Journals (Sweden)

V. V. Galchenko

2010-12-01

Full Text Available In present paper the burnup credit calculations for TK-8 transport container and SVJP-1 spent fuel storage fa-cility of pool type with RBMK-1000 spent fuel during 100-years of cooling time were performed for criticality safety analysis purpose using MCNP and SCALE codes. Only actinides were taken into account for these critical systems. Two approaches were analyzed with isotopes distribution calculations along fuel assembly height and without it. The results show that subcriticality margin is increased considerably using burnup credit and isotopes distribution along fuel assembly height made this value more reasonable.
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
The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)

Energy Technology Data Exchange (ETDEWEB)

Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail: mrgual@ctn.isctn.edu.cu

2000-07-01

The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)
Benchmark calculation of SCALE-PC 4.3 CSAS6 module and burnup credit criticality analysis

Energy Technology Data Exchange (ETDEWEB)

Shin, Hee Sung; Ro, Seong Gy; Shin, Young Joon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-12-01

Calculation biases of SCALE-PC CSAS6 module for PWR spent fuel, metallized spent fuel and solution of nuclear materials have been determined on the basis of the benchmark to be 0.01100, 0.02650 and 0.00997, respectively. With the aid of the code system, nuclear criticality safety analysis for the spent fuel storage pool has been carried out to determine the minimum burnup of spent fuel required for safe storage. The criticality safety analysis is performed using three types of isotopic composition of spent fuel: ORIGEN2-calculated isotopic compositions; the conservative inventory obtained from the multiplication of ORIGEN2-calculated isotopic compositions by isotopic correction factors; the conservative inventory of only U, Pu and {sup 241}Am. The results show that the minimum burnup for three cases are 990,6190 and 7270 MWd/tU, respectively in the case of 5.0 wt% initial enriched spent fuel. (author). 74 refs., 68 figs., 35 tabs.
Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO2 and MOX spent fuels

International Nuclear Information System (INIS)

Sasahara, Akihiro; Matsumura, Tetsuo; Nicolaou, G.; Betti, M.; Walker, C.T.

1997-01-01

The post irradiation examinations (PIE) were carried out for high burn-up UO 2 spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, a) isotopic analysis, b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)
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.
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)
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.

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
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
Substep methods for burnup calculations with Bateman solutions

International Nuclear Information System (INIS)

Isotalo, A.E.; Aarnio, P.A.

2011-01-01

Highlights: → Bateman solution based depletion requires constant microscopic reaction rates. → Traditionally constant approximation is used for each depletion step. → Here depletion steps are divided to substeps which are solved sequentially. → This allows piecewise constant, rather than constant, approximation for each step. → Discretization errors are almost completely removed with only minor slowdown. - Abstract: When material changes in burnup calculations are solved by evaluating an explicit solution to the Bateman equations with constant microscopic reaction rates, one has to first predict the development of the reaction rates during the step and then further approximate these predictions with their averages in the depletion calculation. Representing the continuously changing reaction rates with their averages results in some error regardless of how accurately their development was predicted. Since neutronics solutions tend to be computationally expensive, steps in typical calculations are long and the resulting discretization errors significant. In this paper we present a simple solution to reducing these errors: the depletion steps are divided to substeps that are solved sequentially, allowing finer discretization of the reaction rates without additional neutronics solutions. This greatly reduces the discretization errors and, at least when combined with Monte Carlo neutronics, causes only minor slowdown as neutronics dominates the total running time.
Nuclear fuel burnup calculation in a Voronezh type reactor; Analiza izgaranja nuklearnog goriva u reaktoru tipa Voronjez

Energy Technology Data Exchange (ETDEWEB)

Matausek, M; Marinkovic, N; Kocic, A [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1977-07-01

In order to summarize and present our abilities to perform a complex computation of the nuclear fuel burn-up, a systematic review of the available methods, algorithms and computer programmes is given in this paper. The computer programmes quoted have all been developed, modified and tested in our department, so that they can be successfully used in the analysis of nuclear power plants from both physics and economic points of view. For a commercially proven nuclear reactor - reactor of the Voronezh type - an illustrative computation of the fuel burn-up is performed. The typical results are presented and discussed. The conclusion concerns the completion of a modular scheme for the fuel burn-up calculation and the fuel cycle analysis (author)
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)
OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

Energy Technology Data Exchange (ETDEWEB)

Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

2000-09-01

The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)
Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO{sub 2} and MOX spent fuels

Energy Technology Data Exchange (ETDEWEB)

Sasahara, Akihiro; Matsumura, Tetsuo [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.; Nicolaou, G.; Betti, M.; Walker, C.T.

1997-03-01

The post irradiation examinations (PIE) were carried out for high burn-up UO{sub 2} spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, (a) isotopic analysis, (b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)
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
Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Noori-Kalkhoran, Omid, E-mail: o_noori@yahoo.com [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Yarizadeh-Beneh, Mehdi [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

2016-08-15

Highlights: • Calculation of control rod worth in term of burn-up. • Calculation of differential and integral control rod worth. • Developing an external couple. • Modification of thermal-hydraulic profiles in calculations. - Abstract: One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron absorbent rods. In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 nuclear reactor. External coupling of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of the core in various days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. An external coupling algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and control rod worth for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.
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)
Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations; Quantifizierung der Rechengenauigkeit von Codesystemen zum Abbrandkredit durch Experimentnachrechnungen

Energy Technology Data Exchange (ETDEWEB)

Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik

2014-06-15

In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.
Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

International Nuclear Information System (INIS)

Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

2009-01-01

In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)
LASER-R a computer code for reactor cell and burnup calculations in neutron transport theory

International Nuclear Information System (INIS)

Cristian, I.; Cirstoiu, B.; Dumitrache, I.; Cepraga, D.

1976-04-01

The LASER-R code is an IBM 370/135 version of the Westinghouse code, LASER, based on the THERMOS and MUFT codes developped by Poncelet. It can be used to perform thermal reactor cell calculations and burnup calculations. The cell exhibits 3-4 concentric areas: fuel, cladding, moderator and scattering ring. Besides directions for use, a short description of the physical model, numerical methods and output is presented
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)
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
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)
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)
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
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

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)
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)
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.
FURNACE-J, 2-D Diffusion Burnup for Fast Reactors from JAERI Fast-Set

International Nuclear Information System (INIS)

Ikawa, Koji

1984-01-01

1 - Nature of physical problem solved: FURNACEJ is a two-dimensional diffusion-burnup code for use in the detailed burnup analysis of fast reactors. The code is an extension code of the FURNACE. There exists no essential difference between FURNACE and FURNACEJ. However, the latter can deal with JAERI-Fast-Set as its cross section library, while the former is designed to use ABBN set. Additionally, in FURNACEJ, group-dependent and -independent transverse buckling of each region can be computed and punched on cards, if desired. This is prepared for users so as to use them as input data for detailed two-dimensional x-y calculations. 2 - Restrictions on the complexity of the problem: Only r-z geometry is available
Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor

Energy Technology Data Exchange (ETDEWEB)

Noori-Kalkhoran, Omid; Ahangari, R. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research school; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Faculty of Engineering

2017-03-15

In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7?has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.
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
Results of calculation of WWER-440 fuel rods (Kol`skaya-3 NPP) at high burnup

Energy Technology Data Exchange (ETDEWEB)

Scheglov, A; Proselkov, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Panin, M; Pitkin, Yu [Kol` skaya NPP, (Russian Federation); Tzibulya, V [AO Mashinostroitelnij Zavod Electrostal (Russian Federation)

1994-12-31

Thermal-physical characteristics of fuel rods of two fuel assemblies which were operated within 5 - 8 and 5 - 9 core fuel loadings of the Unit 3 of the Kol`skaya NPP are calculated. They have achieved deep burnup during 4-year (> 46 Mwd/kg U) and 5-year (> 48 Mwd/kg U) fuel cycle. Fuel assemblies have been unloaded off the reactor and subjected to a post-irradiation testing. PIN-mod2 code originally designed for modelling of WWER fuel rod behaviour in a quasi-steady-state operation is used. The average fuel rod in the fuel assembly and the fuel rod with maximum burnup are selected. The preliminary comparison of the calculation results with those of the post-irradiation examination shows a satisfactory agreement. On the basis of the results obtained in the post-irradiation experiments an improvement of the model for calculation of fission gas release and creep of the cladding is planned. The results of the analysis performed indicate that the fuel rod completely preserves its working ability; fuel temperature does not exceed 1300{sup o} C; fission gas release does not exceed 4%; maximum gas pressure inside the cladding at the end of campaign does not exceed 2 MPa. 2 tabs., 11 figs., 5 refs.
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
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
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
Monte Carlo burnup simulation of the TAKAHAMA-3 benchmark experiment

International Nuclear Information System (INIS)

Dalle, Hugo M.

2009-01-01

High burnup PWR fuel is currently being studied at CDTN/CNEN-MG. Monte Carlo burnup code system MONTEBURNS is used to characterize the neutronic behavior of the fuel. In order to validate the code system and calculation methodology to be used in this study the Japanese Takahama-3 Benchmark was chosen, as it is the single burnup benchmark experimental data set freely available that partially reproduces the conditions of the fuel under evaluation. The burnup of the three PWR fuel rods of the Takahama-3 burnup benchmark was calculated by MONTEBURNS using the simplest infinite fuel pin cell model and also a more complex representation of an infinite heterogeneous fuel pin cells lattice. Calculations results for the mass of most isotopes of Uranium, Neptunium, Plutonium, Americium, Curium and some fission products, commonly used as burnup monitors, were compared with the Post Irradiation Examinations (PIE) values for all the three fuel rods. Results have shown some sensitivity to the MCNP neutron cross-section data libraries, particularly affected by the temperature in which the evaluated nuclear data files were processed. (author)
Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

Directory of Open Access Journals (Sweden)

Gholamzadeh Zohreh

2014-12-01

Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view
Conservative axial burnup distributions for actinide-only burnup credit

International Nuclear Information System (INIS)

Kang, C.; Lancaster, D.

1997-11-01

Unlike the fresh fuel approach, which assumes the initial isotopic compositions for criticality analyses, any burnup credit methodology must address the proper treatment of axial burnup distributions. A straightforward way of treating a given axial burnup distribution is to segment the fuel assembly into multiple meshes and to model each burnup mesh with the corresponding isotopic compositions. Although this approach represents a significant increase in modeling efforts compared to the uniform average burnup approach, it can adequately determine the reactivity effect of the axial burnup distribution. A major consideration is what axial burnup distributions are appropriate for use in light of many possible distributions depending on core operating conditions and histories. This paper summarizes criticality analyses performed to determine conservative axial burnup distributions. The conservative axial burnup distributions presented in this paper are included in the Topical Report on Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel Packages, Revision 1 submitted in May 1997 by the US Department of Energy (DOE) to the US Nuclear Regulatory Commission (NRC). When approved by NRC, the conservative axial burnup distributions may be used to model PWR spent nuclear fuel for the purpose of gaining actinide only burnup credit
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
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.
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
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)
'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)
Nuclide Importance and the Steady-State Burnup Equation

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Nemoto, Atsushi

2000-01-01

Conventional methods for evaluating some characteristic values of nuclides relating to burnup in a given neutron spectrum are reviewed in a mathematically systematic way, and a new method based on the importance theory is proposed. In this method, these characteristic values of a nuclide are equivalent to the importances of the nuclide. By solving the equation adjoint to the steady-state burnup equation with a properly chosen source term, the importances for all nuclides are obtained simultaneously.The fission number importance, net neutron importance, fission neutron importance, and absorbed neutron importance are evaluated and discussed. The net neutron importance is a measure directly estimating neutron economy, and it can be evaluated simply by calculating the fission neutron importance minus the absorbed neutron importance, where only the absorbed neutron importance depends on the fission product. The fission neutron importance and absorbed neutron importance are analyzed separately, and detailed discussions of the fission product effects are given for the absorbed neutron importance
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

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)
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
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
MTR core loading pattern optimization using burnup dependent group constants

Directory of Open Access Journals (Sweden)

Iqbal Masood

2008-01-01

Full Text Available A diffusion theory based MTR fuel management methodology has been developed for finding superior core loading patterns at any stage for MTR systems, keeping track of burnup of individual fuel assemblies throughout their history. It is based on using burnup dependent group constants obtained by the WIMS-D/4 computer code for standard fuel elements and control fuel elements. This methodology has been implemented in a computer program named BFMTR, which carries out detailed five group diffusion theory calculations using the CITATION code as a subroutine. The core-wide spatial flux and power profiles thus obtained are used for calculating the peak-to-average power and flux-ratios along with the available excess reactivity of the system. The fuel manager can use the BFMTR code for loading pattern optimization for maximizing the excess reactivity, keeping the peak-to-average power as well as flux-ratio within constraints. The results obtained by the BFMTR code have been found to be in good agreement with the corresponding experimental values for the equilibrium core of the Pakistan Research Reactor-1.
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
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
Comparison of burnup calculation results using several evaluated nuclear data files

International Nuclear Information System (INIS)

Suyama, Kenya; Katakura, Jun-ichi; Nomura, Yasushi

2002-01-01

Burn-up calculation and comparison of the results were carried out to clarify the differences among the following latest evaluated nuclear data libraries: JENDL-3.2, ENDF/B-VI and JEF-2.2. The analyses showed that the differences seen among the current evaluated nuclear data libraries are small for evaluation of the amounts of many uranium and plutonium isotopes. However, several nuclides important for evaluation of nuclear fuel cycle as 238 Pu, 244 Cm, 149 Sm and 134 Cs showed large differences among used libraries. The chain analyses for the isotopes were conducted and the reasons for the differences were discussed. Based on the discussion, information of important cross section to obtain better agreement with the experimental results for 238 Pu, 244 Cm, 149 Sm and 134 Cs was shown. (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
REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR

International Nuclear Information System (INIS)

Boettcher, W.; Schmidt, E.

1969-01-01

1 - Nature of physical problem solved: REFLOS is a programme for the evaluation of fuel-loading schemes in heavy water moderated reactors. The problems involved in this study are: a) Burn-up calculation for the reactor cell. b) Determination of reactivity behaviour, power distribution, attainable burn-up for both the running-in period and the equilibrium of a 3-dimensional heterogeneous reactor model; investigation of radial fuel movement schemes. c) Evaluation of mass flows of heavy atoms through the reactor and fuel cycle costs for the running-in, the equilibrium, and the shut down of a power reactor. If the subroutine for treating the reactor cell were replaced by a suitable routine, other reactors with weakly absorbing moderators could be analyzed. 2 - Method of solution: Nuclear constants and isotopic compositions of the different fuels in the reactor are calculated by the cell-burn-up programme and tabulated as functions of the burn-up rate (MWD/T). Starting from a known state of the reactor, the 3-dimensional heterogeneous reactor programme (applying an extension of the technique of Feinberg and Galanin) calculates reactivity and neutron flux distribution using one thermal and one or two fast neutron groups. After a given irradiation time, the new state of the reactor is determined, and new nuclear constants are assigned to the various defined locations in the reactor. Reloading of fuel may occur if the prescribed life of the reactor is reached or if the effective multiplication factor or the power form factor falls below a specified level. The scheme of reloading to be carried out is specified by a load vector, giving the number of channels to be discharged, the kind of movement from one to another channel and the type of fresh fuel to be charged for each single reloading event. After having determined the core states characterizing the equilibrium period, and having decided the fuel reloading scheme for the running-in period of the reactor life, the fuel
OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN

International Nuclear Information System (INIS)

Hesse, Ulrich; Sieberer, Johann

2006-01-01

1 - Description of program or function: In OREST, the 1-dimensional lattice code HAMMER and the isotope generation and depletion code ORIGEN are directly coupled for burnup simulation in light-water reactor fuels (GRS recommended). Additionally heavy water and graphite moderated systems can be calculated. New version differs from the previous version in the following features: An 84-group-library LIB84 for up to 200 isotopes is used to update the 3-group -POISON-XS. LIB84 uses the same energy boundaries as THERMOS and HAMLET in . In this way, high flexibility is achieved in very different reactor models. The coupling factor between THERMOS and HAMLET is now directly transferred from HAMMER to THERES and omits the equation 4 (see page 6 of the manual). Sandwich-reactor fuel reactivity and burnup calculations can be started with NGEOM = 1. Thorium graphite reactivity and burnup calculations can be started with NLIBE = 1. High enriched U-235 heavy water moderated reactivity and burnup calculations can be started. HAMLET libraries in for U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-242, Am-241, Am-243 and Zirconium are updated using resonance parameters. NEA-1324/04: A new version of the module hamme97.f has replaced the old one. 2 - Method of solution: For the user-defined irradiation history, an input data processor generates program loops over small burnup steps for the main codes HAMMER and ORIGEN. The user defined assembly description is transformed to an equivalent HAMMER fuel cell. HAMMER solves the integral neutron transport equation in a four-region cylindrical or sandwiched model with reflecting boundaries and runs with fuel power calculated rod temperatures. ORIGEN runs with HAMMER-calculated cross sections and neutron spectra and calculates isotope concentrations during burnup by solving the buildup-, depletion- and decay-chain equations. An output data processor samples the outputs of the program modules and generates tabular works for the
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.
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
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)
Simulation of High Burnup Structure in UO2 Using Potts Model

International Nuclear Information System (INIS)

Oh, Jae Yong; Koo, Yang Hyun; Lee, Byung Ho

2009-01-01

The evolution of a high burnup structure (HBS) in a light water reactor (LWR) UO 2 fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the UO 2 matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels
Modeling of burnup express-estimation for UO{sub 2}-fuel

Energy Technology Data Exchange (ETDEWEB)

Likhanskii, Vladimir V.; Tokarev, Sergey A.; Vilkhivskaya, Olga V., E-mail: vilhivskaya_olga@mail.ru

2017-03-15

Highlights: • Proposed engineering model estimates fuel burnup by {sup 134}Cs/{sup 137}Cs activity ratio. • Buildup of cesium isotopes relies on changing neutron spectrum in the core cycle. • {sup 134}Cs/{sup 137}Cs activity ratios in FAs with Gd-doped fuel rods are analyzed. • Comparison of the model calculations with the NPPs spike measurements is presented. - Abstract: The paper presents the developed engineering model of cesium isotopes production as function of UO{sub 2}-fuel burnup and an assessment of their activity ratios. The model considers the evolution of linear power of gadolinium-doped fuel rods and fuel rods surrounding them in fuel assemblies with high enrichment fuel, harder neutron spectrum, and the changes in cross-sections of neutron reactions in thermal and epithermal energy areas. Parametrical dependences in the model are based on the fuel operation data for nuclear power plants and on the detailed neutronic-physical calculations of the core. Presented are the results of the model calculations for the {sup 134}Cs/{sup 137}Cs activity ratios in fuel taking into account the parameter of hardness of the neutron spectrum during the first irradiation cycle for fuel with enrichment ranging from 3.6 wt% in {sup 235}U.
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.)
Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations

International Nuclear Information System (INIS)

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

1999-01-01

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

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
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
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)
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)
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.)
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.
Computation of classical triton burnup with high plasma temperature and current

International Nuclear Information System (INIS)

Batistoni, P.

1990-09-01

For comparison with experiment, the expected production of 14-MeV neutrons from the burnup of tritons produced in the d(d,t)p reaction must be computed. An effort was undertaken to compare in detail the computer codes used for this purpose at TFTR and JET. The calculation of the confined fraction of tritons by the different codes agrees to within a few percent. The high electron temperature in the experiments has raised the critical energy of the tritons that are slowing down to near or above the peak of the D-T reactivity, making the ion drag terms more important. When the different codes use the same slowing down formulas, the calculated burnup was within 6% for a case where orbit effects are expected to be small. Then results from codes with and without the effects of finite radial orbit excursions were compared for two test cases. For medium to high current discharges the finite radius effects are only of order 10%. A new version of the TFTR burnup code using an implicit Fokker-Planck solution was written to include the effects of energy diffusion and charge exchange. These effects change the time-integrated yields by only a few percent, but can significantly affect the instantaneous rates in time. Significant populations of hot ions can affect the fusion reactivity, and this effect was also studied. In particular, the d(d,p)t rate can be 10%--15% less than the d(d, 3 He)n rate which is usually used as a direct monitor of the triton source. Finally, a finite particle confinement time for the thermalized tritons can increase the apparent ''burn-up'' either if there is a high thermal deuteron temperature or if there exists a significant beam deuteron density
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
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
Summary report on the international comparison of NEACRP burnup benchmark calculations for high conversion light water reactor lattices

International Nuclear Information System (INIS)

Akie, Hiroshi; Ishiguro, Yukio; Takano, Hideki

1988-10-01

The results of the NEACRP HCLWR cell burnup benchmark calculations are summarized in this report. Fifteen organizations from eight countries participated in this benchmark and submitted twenty solutions. Large differences are still observed among the calculated values of void reactivities and conversion ratios. These differences are mainly caused from the discrepancies in the reaction rates of U-238, Pu-239 and fission products. The physics problems related to these results are briefly investigated in the report. In the specialists' meeting on this benchmark calculations held in April 1988, it was recommended to perform continuous energy Monte Carlo calculations in order to obtain reference solutions for design codes. The conclusions resulted from the specialists' meeting are also presented. (author)
Computer modelling of the WWER fuel elements under high burnup conditions by the computer codes PIN-W and RODQ2D

Energy Technology Data Exchange (ETDEWEB)

Valach, M; Zymak, J; Svoboda, R [Nuclear Research Inst. Rez plc, Rez (Czech Republic)

1997-08-01

This paper presents the development status of the computer codes for the WWER fuel elements thermomechanical behavior modelling under high burnup conditions at the Nuclear Research Institute Rez. The accent is given on the analysis of the results from the parametric calculations, performed by the programmes PIN-W and RODQ2D, rather than on their detailed theoretical description. Several new optional correlations for the UO2 thermal conductivity with degradation effect caused by burnup were implemented into the both codes. Examples of performed calculations document differences between previous and new versions of both programmes. Some recommendations for further development of the codes are given in conclusion. (author). 6 refs, 9 figs.
Computer modelling of the WWER fuel elements under high burnup conditions by the computer codes PIN-W and RODQ2D

International Nuclear Information System (INIS)

Valach, M.; Zymak, J.; Svoboda, R.

1997-01-01

This paper presents the development status of the computer codes for the WWER fuel elements thermomechanical behavior modelling under high burnup conditions at the Nuclear Research Institute Rez. The accent is given on the analysis of the results from the parametric calculations, performed by the programmes PIN-W and RODQ2D, rather than on their detailed theoretical description. Several new optional correlations for the UO2 thermal conductivity with degradation effect caused by burnup were implemented into the both codes. Examples of performed calculations document differences between previous and new versions of both programmes. Some recommendations for further development of the codes are given in conclusion. (author). 6 refs, 9 figs
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.)
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.
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
Technique for sensitivity analysis of space- and energy-dependent burn-up calculations

International Nuclear Information System (INIS)

Williams, M.L.; White, J.R.

1979-01-01

A practical method is presented for sensitivity analysis of the very complex, space-energy dependent burn-up equations, in which the neutron and nuclide fields are coupled nonlinearly. The adjoint burn-up equations that are given are in a form which can be directly implemented into multi-dimensional depletion codes, such as VENTURE/BURNER. The data sensitivity coefficients can be used to determine the effect of data uncertainties on time-dependent depletion responses. Initial condition sensitivity coefficients provide a very effective method for computing the change in end of cycle parameters (such as k/sub eff/, fissile inventory, etc.) due to changes in nuclide concentrations at beginning of cycle
SWAT3.1 - the integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP

International Nuclear Information System (INIS)

Suyama, Kenya; Mochizuki, Hiroki; Takada, Tomoyuki; Ryufuku, Susumu; Okuno, Hiroshi; Murazaki, Minoru; Ohkubo, Kiyoshi

2009-05-01

Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC,which is widely used in Japan, and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinides and the fission products in the spent nuclear fuel. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP, and ORIGEN2. This enables us to treat the arbitrary fuel geometry and to generate the effective cross section data to be used in the burnup calculation with few approximations. This report describes the outline, input data instruction and several examples of the calculation. (author)
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.
Classification of criticality calculations with correlation coefficient method and its application to OECD/NEA burnup credit benchmarks phase III-A and II-A

International Nuclear Information System (INIS)

Okuno, Hiroshi

2003-01-01

A method for classifying benchmark results of criticality calculations according to similarity was proposed in this paper. After formulation of the method utilizing correlation coefficients, it was applied to burnup credit criticality benchmarks Phase III-A and II-A, which were conducted by the Expert Group on Burnup Credit Criticality Safety under auspices of the Nuclear Energy Agency of the Organisation for Economic Cooperation and Development (OECD/NEA). Phase III-A benchmark was a series of criticality calculations for irradiated Boiling Water Reactor (BWR) fuel assemblies, whereas Phase II-A benchmark was a suite of criticality calculations for irradiated Pressurized Water Reactor (PWR) fuel pins. These benchmark problems and their results were summarized. The correlation coefficients were calculated and sets of benchmark calculation results were classified according to the criterion that the values of the correlation coefficients were no less than 0.15 for Phase III-A and 0.10 for Phase II-A benchmarks. When a couple of benchmark calculation results belonged to the same group, one calculation result was found predictable from the other. An example was shown for each of the Benchmarks. While the evaluated nuclear data seemed the main factor for the classification, further investigations were required for finding other factors. (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
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
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)
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
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.
Estimate of fuel burnup spatial a multipurpose reactor in computer simulation

International Nuclear Information System (INIS)

Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes

2015-01-01

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)
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.)
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)
Burn-up credit in criticality safety of PWR spent fuel

Energy Technology Data Exchange (ETDEWEB)

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

2014-12-15

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

International Nuclear Information System (INIS)

Schutte, C.K.

1991-05-01

The burn-up software system MOP is a research tool primary intended to study the behaviour of fission products in any reactor composition. Input data are multi-group cross-sections and data concerning the nuclide chains. An option is available to calculate a fundamental mode neutron spectrum for the specified reactor composition. A separate program can test the consistency of the specified nuclide chains. Options are available to calculate time-dependent cross-sections of lumped fission products and to take account of the leakage of gaseous fission products from the reactor core. The system is written in FORTRAN77 for a CYBER computer, using the operating system NOS/BE. The report gives a detailed technical description of the applied algorithms and the flow and storage of data. Information is provided for adapting the system to other computer configurations. (author). 5 refs.; 11 figs
Establishing a PWR burn-up library

International Nuclear Information System (INIS)

Lutz, D.C.

1981-01-01

Starting out from data file ENDF/B IV /1/, a cross-section library has been established for the calculation of operating conditions in pressurized water reactors of the type used in BIBLIS B. The library includes macroscopic, homogenized 2-group cross-sections for all types of fuel elements used in this reactor, including those equipped with boron glass rods. For their calculation the previous irradiation of the fuel has been taken into consideration by approximation. Information on fuel consumption from cell burn-up calculations has been stored in a separate data file. It was designed as a base for the determination of cross sections to be used in the calculation of the incident ''main-steam pipe fracture''. For this library the description of cross sections as a function of the moderator status chose the water densities at 300 0 C/155 bar, 190 0 C/140 bar and 100 0 C/100 bar as fixed values. The burn-up library has been tested by a three-dimensional calculation for the 1sup(st) cycle of the BIBLIS B-reactor using program QUABOX /2/. This showed variances with the anticipated course concerning critically, which can be explained almost quantitatively by known deficiencies of the ENDF/b-IV library. (orig.) [de
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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-15

A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.
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.
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
WWER-1000 Burnup Credit Benchmark (CB5)

International Nuclear Information System (INIS)

Manolova, M.A.

2002-01-01

In the paper the specification of WWER-1000 Burnup Credit Benchmark first phase (depletion calculations), given. The second phase - criticality calculations for the WWER-1000 fuel pin cell, will be given after the evaluation of the results, obtained at the first phase. The proposed benchmark is a continuation of the WWER benchmark activities in this field (Author)
Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

International Nuclear Information System (INIS)

Wagner, J.C.

2001-01-01

This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses
Development of a Fully-Automated Monte Carlo Burnup Code Monteburns

International Nuclear Information System (INIS)

Poston, D.I.; Trellue, H.R.

1999-01-01

Several computer codes have been developed to perform nuclear burnup calculations over the past few decades. In addition, because of advances in computer technology, it recently has become more desirable to use Monte Carlo techniques for such problems. Monte Carlo techniques generally offer two distinct advantages over discrete ordinate methods: (1) the use of continuous energy cross sections and (2) the ability to model detailed, complex, three-dimensional (3-D) geometries. These advantages allow more accurate burnup results to be obtained, provided that the user possesses the required computing power (which is required for discrete ordinate methods as well). Several linkage codes have been written that combine a Monte Carlo N-particle transport code (such as MCNP TM ) with a radioactive decay and burnup code. This paper describes one such code that was written at Los Alamos National Laboratory: monteburns. Monteburns links MCNP with the isotope generation and depletion code ORIGEN2. The basis for the development of monteburns was the need for a fully automated code that could perform accurate burnup (and other) calculations for any 3-D system (accelerator-driven or a full reactor core). Before the initial development of monteburns, a list of desired attributes was made and is given below. o The code should be fully automated (that is, after the input is set up, no further user interaction is required). . The code should allow for the irradiation of several materials concurrently (each material is evaluated collectively in MCNP and burned separately in 0RIGEN2). o The code should allow the transfer of materials (shuffling) between regions in MCNP. . The code should allow any materials to be added or removed before, during, or after each step in an automated fashion. . The code should not require the user to provide input for 0RIGEN2 and should have minimal MCNP input file requirements (other than a working MCNP deck). . The code should be relatively easy to use
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...
Burn-up calculations for a thorium HTR with one and with two types of fuel particle

Energy Technology Data Exchange (ETDEWEB)

Griggs, C. F.

1975-06-15

Cell burn-up calculations have been made on a thorium pin-cell operating with one or with two types of particle. With one particle, the input thorium and uranium are mixed prior to irradiation and all discharged uranium is recycled. With two particles, the fuel is kept in two streams and only the uranium generated from thorium is recycled. The two models are found to give similar power generations from a given initial U-235 input. The choice between the two types of particle is probably not determined by reactor physics considerations but by the value of the fuel credits and by the cost of fuel fabrication and reprocessing.
Depletion benchmarks calculation of random media using explicit modeling approach of RMC

International Nuclear Information System (INIS)

Liu, Shichang; She, Ding; Liang, Jin-gang; Wang, Kan

2016-01-01

Highlights: • Explicit modeling of RMC is applied to depletion benchmark for HTGR fuel element. • Explicit modeling can provide detailed burnup distribution and burnup heterogeneity. • The results would serve as a supplement for the HTGR fuel depletion benchmark. • The method of adjacent burnup regions combination is proposed for full-core problems. • The combination method can reduce memory footprint, keeping the computing accuracy. - Abstract: Monte Carlo method plays an important role in accurate simulation of random media, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. Three stochastic geometry modeling methods including Random Lattice Method, Chord Length Sampling and explicit modeling approach with mesh acceleration technique, have been implemented in RMC to simulate the particle transport in the dispersed fuels, in which the explicit modeling method is regarded as the best choice. In this paper, the explicit modeling method is applied to the depletion benchmark for HTGR fuel element, and the method of combination of adjacent burnup regions has been proposed and investigated. The results show that the explicit modeling can provide detailed burnup distribution of individual TRISO particles, and this work would serve as a supplement for the HTGR fuel depletion benchmark calculations. The combination of adjacent burnup regions can effectively reduce the memory footprint while keeping the computational accuracy.
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)
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)
Burn-up function of fuel management code for aqueous homogeneous reactors and its validation

International Nuclear Information System (INIS)

Wang Liangzi; Yao Dong; Wang Kan

2011-01-01

Fuel Management Code for Aqueous Homogeneous Reactors (FMCAHR) is developed based on the Monte Carlo transport method, to analyze the physics characteristics of aqueous homogeneous reactors. FMCAHR has the ability of doing resonance treatment, searching for critical rod heights, thermal hydraulic parameters calculation, radiolytic-gas bubbles' calculation and bum-up calculation. This paper introduces the theory model and scheme of its burn-up function, and then compares its calculation results with benchmarks and with DRAGON's burn-up results, which confirms its bum-up computing precision and its applicability in the bum-up calculation and analysis for aqueous solution reactors. (authors)
The measurement of abundance and content of 148Nd monitor for the determination of burnup with mass spectrometry

International Nuclear Information System (INIS)

Gao Shuqin; Li Silin

1992-09-01

The abundance and content of nuclide 148 Nd, which is used as monitor to determine reactor element burnup, were measured by mass spectrometry, and the burnup can be calculated from measured results. The distribution of 148 Nd abundance and content in the axial direction are consistent with the theoretical calculation. The burnup values agree with the data obtained from heavy isotope ratio and radiochemistry methods within the errors of 4.0% and 2.8% respectively
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)
Fission product model for BWR analysis with improved accuracy in high burnup

International Nuclear Information System (INIS)

Ikehara, Tadashi; Yamamoto, Munenari; Ando, Yoshihira

1998-01-01

A new fission product (FP) chain model has been studied to be used in a BWR lattice calculation. In attempting to establish the model, two requirements, i.e. the accuracy in predicting burnup reactivity and the easiness in practical application, are simultaneously considered. The resultant FP model consists of 81 explicit FP nuclides and two lumped pseudo nuclides having the absorption cross sections independent of burnup history and fuel composition. For the verification, extensive numerical tests covering over a wide range of operational conditions and fuel compositions have been carried out. The results indicate that the estimated errors in burnup reactivity are within 0.1%Δk for exposures up to 100GWd/t. It is concluded that the present model can offer a high degree of accuracy for FP representation in BWR lattice calculation. (author)
Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

Energy Technology Data Exchange (ETDEWEB)

Schneider, E. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, 1 Univ. Place C2200, Austin, TX 78712 (United States); Deinert, M.; Bingham Cady, K. [Dept. of Theoretical and Applied Mechanics, Cornell Univ., Ithaca, NY 14853 (United States)

2006-07-01

Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)
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.
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.)

Accuracy assessment of a new Monte Carlo based burnup computer code

International Nuclear Information System (INIS)

El Bakkari, B.; ElBardouni, T.; Nacir, B.; ElYounoussi, C.; Boulaich, Y.; Meroun, O.; Zoubair, M.; Chakir, E.

2012-01-01

Highlights: ► A new burnup code called BUCAL1 was developed. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► Validation of BUCAL1 was done by code to code comparison using VVER-1000 LEU Benchmark Assembly. ► Differences from BM value were found to be ± 600 pcm for k ∞ and ±6% for the isotopic compositions. ► The effect on reactivity due to the burnup of Gd isotopes is well reproduced by BUCAL1. - Abstract: This study aims to test for the suitability and accuracy of a new home-made Monte Carlo burnup code, called BUCAL1, by investigating and predicting the neutronic behavior of a “VVER-1000 LEU Assembly Computational Benchmark”, at lattice level. BUCAL1 uses MCNP tally information directly in the computation; this approach allows performing straightforward and accurate calculation without having to use the calculated group fluxes to perform transmutation analysis in a separate code. ENDF/B-VII evaluated nuclear data library was used in these calculations. Processing of the data library is performed using recent updates of NJOY99 system. Code to code comparisons with the reported Nuclear OECD/NEA results are presented and analyzed.
Chemical analyses and calculation of isotopic compositions of high-burnup UO{sub 2} fuels and MOX fuels

Energy Technology Data Exchange (ETDEWEB)

Matsumura, Tetsuo; Sasahara, Akihiro [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

2001-08-01

Chemical analysis activities of isotopic compositions of high-burnup UO{sub 2} fuels and MOX fuels in CRIEPI and calculation evaluation are reviewed briefly. C/E values of ORIGEN2, in which original libraries and JENDL-3.2 libraries are used, and other codes with chemical analysis data are reviewed and evaluated. Isotopic compositions of main U and Pu in fuels can be evaluated within 10% relative errors by suitable libraries and codes. Void ratio is effective parameter for C/E values in BWR fuels. JENDL-3.2 library shows remarkable improvement compared with original libraries in isotopic composition evaluations of FP nuclides. (author)
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)
Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

International Nuclear Information System (INIS)

El bakkari, B.; El Bardouni, T.; Merroun, O.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Chakir, E.

2009-01-01

The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc...). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called 'BUCAL1'. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k ∞ ) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.
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
Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

Science.gov (United States)

Muratov, V. G.; Lopatkin, A. V.

An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.
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)
Application of Integral Ex-Core and Differential In-Core Neutron Measurements for Adjustment of Fuel Burn-Up Distributions in VVER-1000

Science.gov (United States)

Borodkin, Pavel G.; Borodkin, Gennady I.; Khrennikov, Nikolay N.

2010-10-01

The paper deals with calculational and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Time-integrated neutron source distributions used for DORT calculations were prepared by two different approaches based on a) calculated fuel burn-up (standard routine procedure) and b) in-core measurements by means of SPD & TC (new approach). Taking into account that fuel burn-up distributions in operating VVER may be evaluated now by analytical methods (calculations) only it is needed to develop new approaches for testing and correction of calculational evaluations. Results presented in this paper allow to consider a reverse task of alternative estimation of fuel burn-up distributions. The approach proposed is based on adjustment (fitting) of time-integrated neutron source distributions, and hence fuel burn-up patterns in some part of reactor core, on the base of ex-core neutron leakage measurement, neutron-physical calculation and in-core SPD & TC measurement data.
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
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
Parametric studies of the effect of MOx environment and control rods for PWR-UOx burnup credit implementation

International Nuclear Information System (INIS)

Barreau, Anne; Roque, Benedicte; Marimbeau, Pierre; Venard, Christophe; Bioux, Philippe; Toubon, Herve

2003-01-01

The increase of PWR-UOX fuel initial enrichment and the extensive needs for spent fuel storage or cask capacities reinforce the interest in taking burnup credit into account in criticality calculations. However, this utilization of credit for fuel burnup requires the definition of a methodology that ensures the conservatism of calculations. In order to guarantee the conservatism of the spent fuel inventory calculation, a depletion calculation scheme for burnup credit is under development. This paper presents the studies on the main parameters which have an effect on nuclides concentration: the presence of control rods during depletion and the fuel assembly environment, particularly the presence of MOx fuels around the UO 2 assembly. Reactivity effects which are relevant to these parameters are then presented, and physics phenomena are identified. (author)
Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

Directory of Open Access Journals (Sweden)

Muhammad Atta

2011-01-01

Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.
Burn-up TRIGA Mark II benchmark experiment

International Nuclear Information System (INIS)

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

1998-01-01

Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(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
Biases and statistical errors in Monte Carlo burnup calculations: an unbiased stochastic scheme to solve Boltzmann/Bateman coupled equations

International Nuclear Information System (INIS)

Dumonteil, E.; Diop, C.M.

2011-01-01

External linking scripts between Monte Carlo transport codes and burnup codes, and complete integration of burnup capability into Monte Carlo transport codes, have been or are currently being developed. Monte Carlo linked burnup methodologies may serve as an excellent benchmark for new deterministic burnup codes used for advanced systems; however, there are some instances where deterministic methodologies break down (i.e., heavily angularly biased systems containing exotic materials without proper group structure) and Monte Carlo burn up may serve as an actual design tool. Therefore, researchers are also developing these capabilities in order to examine complex, three-dimensional exotic material systems that do not contain benchmark data. Providing a reference scheme implies being able to associate statistical errors to any neutronic value of interest like k(eff), reaction rates, fluxes, etc. Usually in Monte Carlo, standard deviations are associated with a particular value by performing different independent and identical simulations (also referred to as 'cycles', 'batches', or 'replicas'), but this is only valid if the calculation itself is not biased. And, as will be shown in this paper, there is a bias in the methodology that consists of coupling transport and depletion codes because Bateman equations are not linear functions of the fluxes or of the reaction rates (those quantities being always measured with an uncertainty). Therefore, we have to quantify and correct this bias. This will be achieved by deriving an unbiased minimum variance estimator of a matrix exponential function of a normal mean. The result is then used to propose a reference scheme to solve Boltzmann/Bateman coupled equations, thanks to Monte Carlo transport codes. Numerical tests will be performed with an ad hoc Monte Carlo code on a very simple depletion case and will be compared to the theoretical results obtained with the reference scheme. Finally, the statistical error propagation
Fuel management and core design code systems for pressurized water reactor neutronic calculations

International Nuclear Information System (INIS)

Ahnert, C.; Arayones, J.M.

1985-01-01

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

Energy Technology Data Exchange (ETDEWEB)

Ramirez S, J. R.; Pineda A, R.; Martinez C, E.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2017-09-15

The burnup of actinides in a nuclear reactor is been proposed as part of an advanced nuclear fuel cycle, this process would close the fuel cycle recycling some of the radioactive material produced in the open nuclear fuel cycle. These actinides are found in the spent nuclear fuel from nuclear power reactors at the end of their burnup in the reactor. Previous studies of actinides recycling in thermal reactors show that would be possible reduce the amounts of actinides at least in 50% of the recycled amounts. in this work, the amounts of actinides that can be burned in a fast reactor is calculated, very interesting results surge from the calculations, first, the amounts of actinides generated by the fuel is higher than for thermal fuel and the composition of the actinides vector is different as in fuel for thermal reactor the main isotope is the {sup 237}Np in the fuel for fast reactor the main isotope is the {sup 241}Am, finally it is concluded that the fast reactor, also generates important amounts of waste. (Author)
Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

Energy Technology Data Exchange (ETDEWEB)

Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

2010-04-01

In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU
Application of burnup credit for PWR spent fuel storage pool

International Nuclear Information System (INIS)

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

1999-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-02-01

Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)
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.
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
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.
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
Analysis of burnup and isotopic compositions of BWR 9 x 9 UO2 fuel assemblies

International Nuclear Information System (INIS)

Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T.

2012-01-01

In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO 2 fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for 238 Pu, 144 Nd, 145 Nd, 146 Nd, 148 Nd, 134 Cs, 154 Eu, 152 Sm, 154 Gd, and 157 Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)
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.
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)
Evaluation of burnup characteristics and energy deposition during NSRR pulse irradiation tests on irradiated BWR fuels

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio

2000-11-01

Pulse irradiation tests of irradiated fuel are performed in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under Reactivity Initiated Accident Conditions (RIA). The severity of the RIA is represented by energy deposition or peak fuel enthalpy during the power excursion. In case of the irradiated fuel tests, the energy deposition varies depending both on the amounts and distribution of residual fissile and neutron absorbing fission products generated during the base irradiation. Thus, proper fuel burnup characterization, especially for low enriched commercial fuels, is important, because plutonium (Pu) takes a large part of fissile and its generation depends on the neutron spectrum during the base irradiation. Fuel burnup calculations were conducted with ORIGEN2, RODBURN and SWAT codes for the BWR fuels tested in the NSRR. The calculation results were compared with the measured isotope concentrations and used for the NSRR neutron calculations to evaluate energy depositions of the test fuel. The comparison of the code calculations and the measurements revealed that the neutron spectrum change due to difference in void fraction altered Pu generation and energy deposition in the NSRR tests considerably. With the properly evaluated neutron spectrum, the combined burnup and NSRR neutron calculation gave reasonably good evaluation of the energy deposition. The calculations provided radial distributions of the fission product accumulation during the base irradiation and power distribution during the NSRR pulse irradiation, which were important for the evaluation of both burnup characteristics and fission gas release behavior. (author)
Implementation of burnup in FERM nodal computer code

International Nuclear Information System (INIS)

Yoriyaz, H.; Nakata, H.

1986-01-01

In this work a spatial burnup scheme and feedback effects has been implemented into the FERM [1] ('Finite Element Response Matrix') program. The spatially dependent neutronic parameters have been considered in three levels: zonewise calculation, assemblywise calculation and pointwise calculation. The results have been compared with the results obtained by CITATION [2] program and showed that the processing time in the FERM code has been hundred of times shorter and no significant difference has been observed in the assembly average power distribution. (Author) [pt
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
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.)
Critical assessment of the pore size distribution in the rim region of high burnup UO_2 fuels

International Nuclear Information System (INIS)

Cappia, F.; Pizzocri, D.; Schubert, A.; Van Uffelen, P.; Paperini, G.; Pellottiero, D.; Macián-Juan, R.; Rondinella, V.V.

2016-01-01

A new methodology is introduced to analyse porosity data in the high burnup structure. Image analysis is coupled with the adaptive kernel density estimator to obtain a detailed characterisation of the pore size distribution, without a-priori assumption on the functional form of the distribution. Subsequently, stereological analysis is carried out. The method shows advantages compared to the classical approach based on the histogram in terms of detail in the description and accuracy within the experimental limits. Results are compared to the approximation of a log-normal distribution. In the investigated local burnup range (80–200 GWd/tHM), the agreement of the two approaches is satisfactory. From the obtained total pore density and mean pore diameter as a function of local burnup, pore coarsening is observed starting from ≈100 GWd/tHM, in agreement with a previous investigation. - Highlights: • A new methodology to analyse porosity is introduced. • The method shows advantages compared to the histogram. • Pore density and mean diameter data vs. burnup are presented. • Pore coarsening is observed starting from ≈100 GWd/tHM.
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
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)
Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

Energy Technology Data Exchange (ETDEWEB)

DeHart, M.D.

1996-05-01

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

International Nuclear Information System (INIS)

DeHart, M.D.

1996-05-01

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

International Nuclear Information System (INIS)

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

1998-01-01

The results and conclusions from a six-year study by an international benchmarking group in the comparison of computational methods for evaluating burnup credit in criticality safety analyses is presented. Approximately 20 participants from 12 countries have provided results for most problems. Four detailed benchmark problems for pressurized-water-reactor fuel have been completed. Results from work being finalized, addressing burnup credit for boiling-water-reactor fuel, are discussed, as well as planned activities for additional benchmarks, including mixed-oxide fuels, and other activities

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
Optimalisation Of Oxide Burn-Up Enhanced For RSG-Gas Core

International Nuclear Information System (INIS)

Tukiran; Sembiring, Tagor Malem

2000-01-01

Strategy of fuel management of the RSG-Gas core has been changed from 6/1 to 5/1 pattern so the evaluation of fuel management is necessary to be done. The aim of evaluation is to look for the optimal fuel management so that the fuel can be stayed longer in the core and finally can save cost of operation. Using Batan-EQUIL-2D code did the evaluation of fuel management with 5/1 pattern. The result of evaluation is used to choose which one is more advantage without break the safety margin which is available in the Safety Analysis Report (SAR) firstly, the fuel management was calculated with core excess reactivity of 9,2% criteria. Secondly, fuel burn-up maximum of 56% criteria and the last, fuel burn-up maximum of 64% criteria. From the result of fuel management calculation of the RSG-Gas equilibrium core can be concluded that the optimal RSG-Gas equilibrium core with 5/1 pattern is if the fuel burn-up maximum 64% and the energy in a cycle of operation is 715 MWD. The fuel can be added one more step in the core without break any safety margin. It means that the RSG-Gas equilibrium core can save fuel and cost reduction
Neutronic calculations of PARR-1 cores using LEU silicide fuel

International Nuclear Information System (INIS)

Arshad, M.; Bakhtyar, S.; Hayat, T.; Salahuddin, A.

1991-08-01

Detailed neutronic calculations have been carried out for different PARR-1 cores utilizing low enriched uranium (LEU) silicide fuel and operating at an upgraded power of 9 MW. The calculations include the search for critical loadings in open and stall ends of the pool, neutronic analysis of the first full equilibrium core and calculations cores. The burnup study of inventory have also been carried out. Further, the reactivity coefficients of the first full power operation core are evaluated for use in the accident analysis. 14 figs. (author)
Research on burnup physics

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1974-07-01

One of the major problems in burnup studies is the reasonably fast and accurate calculation of the space-and-energy dependent neutron flux and reaction rates for realistic power reactor fuel geometries and compositions, and its optimal integration in the global reactor calculations. The scope of the present research was to develop improved methods trying to satisfy the above requirements. In the epithermal region, simple and efficient approximation is proposed which allows the analytical solution for the space dependence of the spherical harmonics flux moments, and hence the derivation of the recurrence relations between he flux moments at successive lethargy pivotal points. A new matrix formalism to invert the coefficient matrix of band structure resulted in a reduce computer time and memory demands. The research on epithermal region is finalized in computing programme SPLET, which calculates the space-lethargy distribution of the spherical harmonics neutron flux moments, and the related integral quantities as reaction rates and resonance integrals. For partial verification of the above methods a Monte Carlo procedure was developed. Using point-wise representation of variables, a flexible and fast convergent integral transport method SEPT i developed. Expanding the neutron source and flux in finite series of arbitrary polynomials, the space-and-energy dependent integral transport equation is transformed into a general linear algebraic form, which is solved numerically. A simple and efficient procedure for deriving multipoint equations and constructing matrix is proposed and examined, and no unwanted oscillations were noticed. The energy point method was combined with the spherical harmonics method as well. A multi zone few-group program SPECTAR for global reactor calculations was developed. For testing, the flux distribution, neutron leakage and effective multiplication factor for the PWR reactor of the power station San Onofre were calculated. In order to verify
Calculation and analysis of burnup and optimum core design in accelerator driven sub-critical system

International Nuclear Information System (INIS)

Wang Yuwei; Yang Yongwei; Cui Pengfei

2011-01-01

The premise of the accelerator driven sub-critical system (ADS) in the accident is still subcritical, the biggest k eff change with burn time is less than 1.5% and the cladding material, HT9 steel, can withstand the maximum radiation damage, core fuel area is divided into fuel transmutation area and fuel multiplication area, and fuel transmutation area maintains the same fuel composition in the whole process. Through the analysis of the composition of the fuel, shape of core layout and the power distribution, etc., supposed outer and inner Pu enrichment ratio range of 1.0-1.5, then the fuel components of fuel multiplication area was adjusted. Time evolution of k eff was calculated by COUPLED2 which coupled with MCNP and ORIGEN. At the same time the power peaking factors, minoractinides transmutation rate desired to maximization and burnup were considered. A sub-critical system fitting for engineering practice was established. (authors)
Critical assessment of the pore size distribution in the rim region of high burnup UO{sub 2} fuels

Energy Technology Data Exchange (ETDEWEB)

Cappia, F. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Department of Nuclear Engineering, Faculty of Mechanical Engineering, Technische Universität München, D-85748 Garching bei München (Germany); Pizzocri, D. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Nuclear Engineering Division, Energy Department, Politecnico di Milano, 20156 Milano (Italy); Schubert, A.; Van Uffelen, P.; Paperini, G.; Pellottiero, D. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Macián-Juan, R. [Department of Nuclear Engineering, Faculty of Mechanical Engineering, Technische Universität München, D-85748 Garching bei München (Germany); Rondinella, V.V., E-mail: Vincenzo.RONDINELLA@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany)

2016-11-15

A new methodology is introduced to analyse porosity data in the high burnup structure. Image analysis is coupled with the adaptive kernel density estimator to obtain a detailed characterisation of the pore size distribution, without a-priori assumption on the functional form of the distribution. Subsequently, stereological analysis is carried out. The method shows advantages compared to the classical approach based on the histogram in terms of detail in the description and accuracy within the experimental limits. Results are compared to the approximation of a log-normal distribution. In the investigated local burnup range (80–200 GWd/tHM), the agreement of the two approaches is satisfactory. From the obtained total pore density and mean pore diameter as a function of local burnup, pore coarsening is observed starting from ≈100 GWd/tHM, in agreement with a previous investigation. - Highlights: • A new methodology to analyse porosity is introduced. • The method shows advantages compared to the histogram. • Pore density and mean diameter data vs. burnup are presented. • Pore coarsening is observed starting from ≈100 GWd/tHM.
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
Evaluation of Isotopic Measurements and Burn-up Value of Sample GU3 of ARIANE Project

Energy Technology Data Exchange (ETDEWEB)

Tore, C.; Rodriguez Rivada, A.

2014-07-01

Estimation of the burn-up value of irradiated fuel and its isotopic composition are important for criticality analysis, spent fuel management and source term estimation. The practical way to estimate the irradiated fuel composition and burn.up value is calculation with validated code and nuclear data. Such validation of the neutronic codes and nuclear data requires the benchmarking with measured values. (Author)
Application of routine methods for the inspector fuel burn-up determination and identification of displacement of spent fuel elements by dummy elements

International Nuclear Information System (INIS)

Rohar, S.

1979-08-01

14 irradiated assemblies were analyzed using nondestructive high resolution gamma spectrometry (HRGS). Measured and calculated (on the basis of calorimetric data) axial burnup profiles and average burnup values were compared. The measurements of spent fuel were performed in the Bohunice A-1 dry hot cell by using a proper collimating system and the standard Agency equipment, consisting of PGT intrinsic Ge detectors and Silena MCA with 1024 channels. The method of 134 Cs/ 137 Cs fission product activity ratio was used for burnup determination. It was found that the burnup values for 14 measured assemblies determined by HRGS were systematically lower than the calculated values with about 4-5%. The difference between the nondestructively determined burnup value of the 2N0053 assembly (average over 11 measured points) and destructively determined burnup (average over 19 measured points) was less than 2%. Passive neutron measurements of the irradiated assembly showed that the neutron counting rate was high enough for practical use and that the neutron and gamma profiles were similar and close to the burnup profile. Some calculations of gamma ray activity angular distribution were made for different numbers of dummy elements inside the irradiated assemblies. The results show that, by using gamma spectrometry transversal method, it is possible to find a significant number of dummy elements in different types of assemblies
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
Influence of high burnup on the decay heat power of spent fuel at long-term storage

International Nuclear Information System (INIS)

Bergelson, B.; Gerasimov, A.; Tikhomirov, G.

2005-01-01

Development and application of advanced fuel with higher burnup is now in practice of NPP with light water reactors in an increasing number of countries. High burnup allows to decrease significantly consumption of uranium. However, spent fuel of this type contains increased amount of high active actinides and fission products in comparison with spent fuel of common-type burnup. Therefore extended time of storage, improved cooling system of the storage facility will be required along with more strong radiation protection during storage, transportation and processing. Calculated data on decay heat power of spent uranium fuel of light water VVER-1000 type reactor are discussed in the paper. Long-term storage of discharged fuel during 100000 years is considered. Calculations were made for burnups of 40-70 MW d/kg. In the initial 50-year period of storage, power of fission products is much higher than that of actinides. Power of gamma-radiation is mainly due to fission products. During subsequent storage power of fission products quickly decreases, the main contribution to the power is given by actinides rather than by fission products. (author)
Characterisation of high-burnup LWR fuel rods through gamma tomography

International Nuclear Information System (INIS)

Caruso, S.

2007-01-01

europium fission-product radionuclides. As indicated above, results provided by the transmission tomography measurements were employed in the emission tomography reconstruction phase, together with a calculated global efficiency matrix and input sinograms derived from the processing of measured projections. Different tomographic algorithms were tested and 'tuned', on the basis of known test distributions, before being applied to the actual fuel rod measurements. Amongst the various possibilities, the Paraboloidal Surrogates Coordinate Ascent penalised likelihood method has been chosen for presentation of the final results, because it ensures high precision, especially in resolving the most difficult peripheral regions of the rods. The results of the emission tomography have indicated large central depressions in the caesium distributions, but of varying extent from sample to sample. Particularly interesting is the case of the 126 GWd/t sample, showing a very deep central depression (a factor of ∼ 2. 5 for 137 Cs, a factor of ∼ 3 for 134 Cs). Differences in the relative activity distributions of 137 Cs and 134 Cs have, in fact, been observed for all the samples. The depression of 134 Cs is more marked than that of 137 Cs, probably due to the different origins of the two isotopes. In contrast, the europium shows an almost flat distribution. In order to support the tomographically measured caesium distributions, the results of destructive chemical techniques applied on samples from the same fuel rod (126 GWd/t sample) were examined and found to show reasonably good agreement with the tomography, thus confirming the depressed distributions at the centre of the rod. In addition to the tomographic reconstructions, the present research has also investigated the possibility to use single isotope activities, and/or isotopic concentration ratios from 134 Cs, 137 Cs and 154 Eu ,as burnup indicators at very high and ultra-high burnups. The corresponding non
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)
The influence of rhodium burn-up on the sensitivity of rhodium self-powered neutron detectors

International Nuclear Information System (INIS)

Erben, O.

1980-01-01

Depression and self-shielding coefficients are presented for thermal and epithermal neutron flux densities. Functions are shown describing the distribution of beta particle sources on the emitter cross section for 0 to 50% rhodium burnup. The values are calculated of detector sensitivity to thermal and epithermal neutron flux densities for the said burnup for main types of rhodium SPN detectors made by SODERN. (J.B.)
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)
Accident Source Terms for Pressurized Water Reactors with High-Burnup Cores Calculated using MELCOR 1.8.5.

Energy Technology Data Exchange (ETDEWEB)

Gauntt, Randall O.; Goldmann, Andrew; Kalinich, Donald A.; Powers, Dana A.

2016-12-01

In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs 2 MoO 4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU
Effects of thermal-hydraulic feedback on burnup modeling of the deep burn modular high temperature reactor (DB-MHR)

International Nuclear Information System (INIS)

Bei, Yea; Wen, Wua; Di, Yuna; Stubbins, J.F.; Venneri, F.

2007-01-01

The Deep-Burn concept investigates the use of commercial high temperature gas-cooled reactors such as modular helium reactors (DB-MHR) to transmute spent fuel from light water reactors (LWRs). An essential feature of this technology is the fabrication of spent fuel into TRISO particles with full transuranic composition to achieve very extensive destruction levels (deep-burn) in a one-pass fuel cycle. Due to the strong temperature influence on the cross sections of transuranics, the coupling between temperature and neutronics is very important to be able to simulate realistic operations of the deep burn reactor. In this study, detailed simulations of the DB-MHR operation are performed with a Monte Carlo code system (MCNP-5 + ORIGEN-2.2 + MONTEBURNS-2 for neutronics calculations), POKE code (General Atomics, for thermohydraulics calculations) and NJOY-99 code (for processing nuclear data libraries), called MHRBURNS. Resulting power densities of fuel blocks (from neutronics calculations) are provided as input to the POKE code, which in turn, calculates new temperature distributions. The temperature distributions obtained from POKE are used to update the MCNP input, and NJOY is called to process new nuclear cross sections based on appropriate temperatures. These steps are repeated to calculate the entire burnup performance of the system. In this preliminary study only the feedback on graphite temperature is taken into account. It is observed that the temperature feedback results show a 200 K higher temperature and thus a slight difference in 237 Np and 239 Pu destruction rates, although the overall burnup rates remain the same
Effect of a time varying power level in EBR-II on mixed-oxide fuel burnup

International Nuclear Information System (INIS)

Stone, I.Z.; Jost, J.W.; Baker, R.B.

1979-01-01

A refined prediction of burnup of mixed-oxide fuel in EBR-2 is compared with measured data. The calculation utilizes a time-varying power factor and results in a general improvement to previous calculations

Russian system of computerized analysis for licensing at atomic industry (SCALA) and its validation on ICSBEP handbook data and some burnup calculations

International Nuclear Information System (INIS)

Ivanova, T.; Nikolaev, M.; Polyakov, A.; Saraeva, T.; Tsiboulia, A.

2000-01-01

The System of Computerized Analysis for Licensing at Atomic industry (SCALA) is a Russian analogue of the well-known SCALE system. For criticality evaluations the ABBN-93 system is used with TWODANT and with joined American KENO and Russian MMK Monte-Carlo code MMKKENO. Using the same cross sections and input models, all these codes give results that coincide within the statistical uncertainties (for Monte-Carlo codes). Validation of criticality calculations using SCALA was performed using data presented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Another task of the work was to test the burnup capability of SCALA system in complex geometry in compare with other codes. Benchmark models of VVER type reactor assemblies with UO 2 and MOX fuel including the cases with burnable gadolinium absorbers were calculated. KENO-VI and MMK codes were used for power distribution calculations, ORIGEN code was used for the isotopic kinetics calculations. (authors)
Establishing the fuel burn-up measuring system for 106 irradiated assemblies of Dalat reactor by using gamma spectrometer method

International Nuclear Information System (INIS)

Nguyen Minh Tuan; Pham Quang Huy; Tran Tri Vien; Trang Cao Su; Tran Quoc Duong; Dang Tran Thai Nguyen

2013-01-01

The fuel burn-up is an important parameter needed to be monitored and determined during a reactor operation and fuel management. The fuel burn-up can be calculated using computer codes and experimentally measured. This work presents the theory and experimental method applied to determine the burn-up of the irradiated and 36% enriched VVR-M2 fuel type assemblies of Dalat reactor. The method is based on measurement of Cs-137 absolute specific activity using gamma spectrometer. Designed measuring system consists of a collimator tube, high purity Germanium detector (HPGe) and associated electronics modules and online computer data acquisition system. The obtained results of measurement are comparable with theoretically calculated results. (author)
Computer programs for TRIGA calibration, burnup evaluation, and bookkeeping

International Nuclear Information System (INIS)

Nelson, George W.

1978-01-01

Several computer programs have been developed at the University of Arizona to assist the direction and operation of the TRIGA Reactor Laboratory. The programs fall into the following three categories: 1. Programs for calculation of burnup of each fuel element in the reactor core, for maintaining an inventory of fuel element location and fissile content at any time, and for evaluation of the reactivity effects of burnup or proposed fuel element rearrangement in the core. 2. Programs for evaluation, function fitting, and tabulation of control rod measurements. 3. Bookkeeping programs to summarize and tabulate reactor runs and irradiations according to time, energy release, purpose, responsible party, etc. These summarized data are reported in an annual operating report for the facility. The use of these programs has saved innumerable hours of repetitious work, assuring more accurate, objective results, and requiring a minimum of effort to repeat calculations when input data are modified. The programs are written in FORTRAN-IV, and have been used on a CDC-6400 computer. (author)
Evaluation of the performance of mini-WIMS in design calculations for SGHWR's

International Nuclear Information System (INIS)

Halsall, M.J.

1980-07-01

In order to use the WIMS code for SGHWR design calculations it is desirable to reduce the computing time to a minimum. To this end, a study has been made of the effects of using condensed data libraries with few groups in the main transport routine and with coarse mesh representations. The results of initial lattice calculations are given in considerable detail for a set of SGHW experimental cores. The effects of condensation on attainable burnup and irradiated fuel composition for natural and enriched power reactor lattices have also been studied. Comparisons between detailed and condensed WIMS calculations are the main theme of the report but METHUSELAH and experimental results are included whenever possible. (author)
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
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
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
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)
Influence of spectral history on PWR full core calculation results

International Nuclear Information System (INIS)

Bilodid, Y.; Mittag, S.

2011-01-01

The few-group cross section libraries, used by reactor dynamics codes, are affected by the spectral history effect-a dependence of fuel cross sections not only on burnup, but also on local spectral conditions during burnup. A cross section correction method based on Pu-239 concentration was implemented in the reactor dynamic code DYN3D. This paper describes the influence of a cross section correction on full-core calculation results. Steady-state and burnup characteristics of a PWR equilibrium cycle, calculated by DYN3D with and without cross section corrections, are compared. A study has shown a significant influence of spectral history on axial power and burnup distributions as well as on calculated cycle length. An impact of the correction on transient calculations is studied for a control rod ejection example. (Authors)
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)
The burn-up credit physics and the 40. Minerve anniversary; La physique du credit Burn-Up et le 40. anniversaire de Minerve

Energy Technology Data Exchange (ETDEWEB)

Santamarina, A [CEA/Cadarache, Departement d' Etudes des Reacteurs, DER/SPRC, 13 - Saint-Paul-lez-Durance (France); Toubon, H [Cogema, 78 - Velizy Villacoublay (France); Trakas, C [FRAMATOME, 92 - Paris La Defense (France); and others

2000-03-21

The technical meeting organized by the SFEN on the burn-up credit (CBU) physics, took place the 23 november 1999 at Cadarache. the first presentation dealt with the economic interest and the neutronic problems of the CBU. Then two papers presented how taking into account the CBU in the industry in matter of transport, storage in pool, reprocessing and criticality calculation (MCNP4/Apollo2-F benchmark). An experimental method for the reactivity measurement through oscillations in the Minerve reactor, has been presented with an analysis of the possible errors. The future research program OSMOSE, taking into account the minor actinides in the CBU, was also developed. The last paper presented the national and international research programs in the CBU domain, in particular experiments realized in CEA/Valduc and the OECD Burn-up Criticality Benchmark Group activities. (A.L.B.)
Burnup characteristics of binary breeder reactors

International Nuclear Information System (INIS)

Dias, A.F.; Nascimento, J.A. do; Ishiguro, Y.

1983-01-01

Burnup calculations of a binary breeder reactor have been done for two cases of fueling. In one case the U 233 /TH fueled inner core and the Pu/U-fueled outer core have the same number of fuel assemblies. In the other case two outermost rings in the inner core are Pu/U-fueled. The second case is considered for an initial phase of thorim cycle introduction when the supply of U 233 could be limited. Results show an efficient breeding on the thorium cycle in both cases. (Author) [pt
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
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)
Present status and future developments of the implementation of burnup credit in spent fuel management systems in Germany

International Nuclear Information System (INIS)

Neuber, J.C.

1998-01-01

The paper describes the experience gained in Germany in applying burnup credit methodologies to wet storage and dry transport systems of spent LWR fuel. It gives a survey of the levels of burnup credit presently used or intended to be used, the regulatory status and future developments planned, the codes used for performing depletion and criticality calculations, the methods applied to verification of these codes, and the methods used to treat parameters specific of burnup credit. In particular it is shown that the effect of axial burnup profiles on wet PWR storage designs based on burnup credit varies from fuel type to fuel type. For wet BWR storage systems the method of estimating a loading curve is described which provides for a given BWR fuel assembly design the minimum required initial burnable absorber content as a function of the initial enrichment of the fuel. (author)
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)
Corrections to the 148Nd method of evaluation of burnup for the PIE samples from Mihama-3 and Genkai-1 reactors

International Nuclear Information System (INIS)

Suyama, Kenya; Mochizuki, Hiroki

2006-01-01

The value of the burnup is one of the most important parameters of samples taken by post-irradiation examination (PIE). Generally, it is evaluated by the Neodymium-148 method. Precise evaluation of the burnup value requires: (1) an effective fission yield of 148 Nd; (2) neutron capture reactions of 147 Nd and 148 Nd; (3) a conversion factor from fissions per initial heavy metal to the burnup unit GWd/t. In this study, the burnup values of the PIE data from Mihama-3 and Genkai-1 PWRs, which were taken by the Japan Atomic Energy Research Institute, were re-evaluated using more accurate corrections for each of these three items. The PIE data were then re-analyzed using SWAT and SWAT2 code systems with JENDL-3.3 library. The re-evaluation of the effective fission yield of 148 Nd has an effect of 1.5-2.0% on burnup values. Considering the neutron capture reactions of 147 Nd and 148 Nd removes dependence of C/E values of 148 Nd on the burnup value. The conversion factor from FIMA(%) to GWd/t changes according to the burnup value. Its effect on the burnup evaluation is small for samples having burnup of larger than 30 GWd/t. The analyses using the corrected burnup values showed that the calculated 148 Nd concentrations and the PIE data is approximately 1%, whereas this was 3-5% in prior analyses. This analysis indicates that the burnup values of samples from Mihama-3 and Genkai-1 PWRs should be corrected by 2-3%. The effect of re-evaluation of the burnup value on the neutron multiplication factor is an approximately 0.6% change in PIE samples having the burnup of larger than 30 GWd/t. Finally, comparison between calculation results using a single pin-cell model and an assembly model is carried out. Because the results agreed with each other within a few percent, we concluded that the single pin-cell model is suitable for the analysis of PIE samples and that the underestimation of plutonium isotopes, which occurred in the previous analyses, does not result from a geometry

Estimation of burnup with cesium isotopes based on gamma-scanning of a instrumented fuel capsule(02F-11K) in hot-cell

International Nuclear Information System (INIS)

Song, Ung Sup; Kim, Hee Moon; Park, Dae Gyu; Paik, Seung Je; Lee, Hong Gi; Choo, Yong Sun; Hong Kwon Pyo

2004-01-01

Many experimental inspection have been performed to obtain the burnup of fuel. In the case, chemical analysis were popular with high reliability. High radioactivity of fuel was severe problem during destructive procedure. Afterward, many researchers have studied calculation of burnup using gamma detector as the non-destructive method. methodologies of gamma-scanning test have been developed as well as higher accuracy of detector. Generally, Cs-137 and Cs-134 are standard isotopes for long-term cooling spent fuel to estimate burnup, because atomic ratio of them follows the linearity with burnup
A validated methodology for evaluating burnup credit in spent fuel casks

International Nuclear Information System (INIS)

Brady, M.C.; Sanders, T.L.

1991-01-01

The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burnup credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various casks geometries, and reactor restart critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias in k eff . Implementation issues affecting licensing requirements and operational procedures are discussed briefly
A validated methodology for evaluating burnup credit in spent fuel casks

International Nuclear Information System (INIS)

Brady, M.C.; Sanders, T.L.

1991-01-01

The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the U.S. Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burnup credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor restart critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias in k eff . Implementation issues affecting licensing requirements and operational procedures are discussed briefly. (Author)
A validated methodology for evaluating burnup credit in spent fuel casks

International Nuclear Information System (INIS)

Brady, M.C.; Sanders, T.L.

1991-01-01

The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burnup credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor restart critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias in k eff . Implementation issues affecting licensing requirements and operational procedures are discussed briefly. 24 refs., 3 tabs
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)
Discrete rod burnup analysis capability in the Westinghouse advanced nodal code

International Nuclear Information System (INIS)

Buechel, R.J.; Fetterman, R.J.; Petrunyak, M.A.

1992-01-01

Core design analysis in the last several years has evolved toward the adoption of nodal-based methods to replace traditional fine-mesh models as the standard neutronic tool for first core and reload design applications throughout the nuclear industry. The accuracy, speed, and reduction in computation requirements associated with the nodal methods have made three-dimensional modeling the preferred approach to obtain the most realistic core model. These methods incorporate detailed rod power reconstruction as well. Certain design applications such as confirmation of fuel rod design limits and fuel reconstitution considerations, for example, require knowledge of the rodwise burnup distribution to avoid unnecessary conservatism in design analyses. The Westinghouse Advanced Nodal Code (ANC) incorporates the capability to generate the intra-assembly pin burnup distribution using an efficient algorithm
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
Calculation of oxygen distribution in uranium-plutonium oxide fuels during irradiation (programme CODIF)

International Nuclear Information System (INIS)

Moreno, A.; Sari, C.

1978-01-01

Radial gradients of oxygen to metal ratio, O/M, in uranium-plutonium oxide fuel pins, during irradiation and at the end of life, have been calculated on the basis of solid-state thermal diffusion using measured values of the heat of transport. A detailed computer model which includes the calculation of temperature profiles and the variation of the average O/M ratio as a function of burn-up is given. Calculations show that oxygen profiles are affected by the isotopic composition of the fuel, by the temperature profiles and by fuel-cladding interactions
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
Advanced fuel cycles and burnup increase of WWER-440 fuel

International Nuclear Information System (INIS)

Proselkov, V.; Saprykin, V.; Scheglov, A.

2003-01-01

Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge
Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library

International Nuclear Information System (INIS)

Chambon, A.; Santamarina, A.; Riffard, C.; Lavaud, F.; Lecarpentier, D.

2013-01-01

Burnup Credit (BUC) is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a 'best estimate' value of the keff, biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 fission products (FPs) of PWR-MOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF- 3.1.1/SHEM library). Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections. (authors)
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
Reactivity management and burn-up management on JRR-3 silicide-fuel-core

International Nuclear Information System (INIS)

Kato, Tomoaki; Araki, Masaaki; Izumo, Hironobu; Kinase, Masami; Torii, Yoshiya; Murayama, Yoji

2007-08-01

On the conversion from uranium-aluminum-dispersion-type fuel (aluminide fuel) to uranium-silicon-aluminum-dispersion-type fuel (silicide fuel), uranium density was increased from 2.2 to 4.8 g/cm 3 with keeping uranium-235 enrichment of 20%. So, burnable absorbers (cadmium wire) were introduced for decreasing excess reactivity caused by the increasing of uranium density. The burnable absorbers influence reactivity during reactor operation. So, the burning of the burnable absorbers was studied and the influence on reactor operation was made cleared. Furthermore, necessary excess reactivity on beginning of operation cycle and the time limit for restart after unplanned reactor shutdown was calculated. On the conversion, limit of fuel burn-up was increased from 50% to 60%. And the fuel exchange procedure was changed from the six-batch dispersion procedure to the fuel burn-up management procedure. The previous estimation of fuel burn-up was required for the planning of fuel exchange, so that the estimation was carried out by means of past operation data. Finally, a new fuel exchange procedure was proposed for effective use of fuel elements. On the procedure, burn-up of spent fuel was defined for each loading position. The average length of fuel's staying in the core can be increased by two percent on the procedure. (author)
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.
CONHOR. Code system for determination of power distribution and burnup for the HOR reactor. Version 1.0.. User's manual

International Nuclear Information System (INIS)

Serov, I.V.; Hoogenboom, J.E.

1993-07-01

The main calculational tool is the CITATION code. CITATION is used for both static and burnup calculations. The pointwise flux density and power distributions obtained from these calculations are used to obtain the values of the desired quantities at the beginning of a burnup cycle. To obtain the most trustful values of the desired quantities CONHOR employs experimental information together with the CITATION calculated flux distributions. Axially averaged foil activation rates are obtained based on both CITATION pointwise flux density distributions and measured foil activity counts. These two sets of activation rates are called the distributions of auxiliary quantities and are compared with each other in order to pick up the corrections to the U-235 number densities in fuel containing elements. The methodical corrections to the calculational auxiliary quantities are obtained on this basis as well. They are used to obtain the methodical corrections to the desired quantities. The corrected desired quantities are the recommended ones. The correction procedure requires the knowledge of the sensitivity coefficients of the average foil activation rates with respect to the U-235 number densities (through the text of this manual U-235 is denoted also and especially in the input-output description sections as a BUrning-COrrected material, or 'BuCo' material). These sensitivity coefficients are calculated by the CONHOR SENS module. CITATION is employed to perform the calculations with perturbed values of U-235 number densities. Burnup calculations can be performed being based on either corrected or uncorrected U-235 number densities. Through the text of this manual XXXX means a 4-symbol identification of the burnup cycle to be studied. XX-1 and XX+1 mean correspondingly the previous and the following cycles. (orig./HP)
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
Present status and future developments of the implementation of burnup credit in spent fuel management systems in Germany

International Nuclear Information System (INIS)

Neuber, J.C.; Kuehl, H.

2001-01-01

This paper describes the experience gained in Germany in implementing burnup credit in wet storage and dry transport systems of spent PWR, BWR, and MOX fuel. It gives a survey of the levels of burnup credit presently used, the regulatory status and activities planned, the fuel depletion codes and criticality calculation codes employed, the verification methods used for validating these codes, the modeling assumptions made to ensure that the burnup credit criticality analysis is based on a fuel irradiation history which leads to bounding neutron multiplication factors, and the implementation of procedures used for fuel loading verification. (author)
Present status and future developments of the implementation of burnup credit in spent fuel management systems in Germany

Energy Technology Data Exchange (ETDEWEB)

Neuber, J C [Siemens Nuclear Power GmbH, Offenbach (Germany); Kuehl, H [Wissenschaftlich-Technische Ingenieurberatung WTI GmbH, Juelich (Germany)

2001-08-01

This paper describes the experience gained in Germany in implementing burnup credit in wet storage and dry transport systems of spent PWR, BWR, and MOX fuel. It gives a survey of the levels of burnup credit presently used, the regulatory status and activities planned, the fuel depletion codes and criticality calculation codes employed, the verification methods used for validating these codes, the modeling assumptions made to ensure that the burnup credit criticality analysis is based on a fuel irradiation history which leads to bounding neutron multiplication factors, and the implementation of procedures used for fuel loading verification. (author)
Corrections to the {sup 148}Nd method of evaluation of burnup for the PIE samples from Mihama-3 and Genkai-1 reactors

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya [Fuel Cycle Facility Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)]. E-mail: suyama.kenya@jaea.go.jp; Mochizuki, Hiroki [Japan Research Institute, Limited, 16 Ichiban-cho, Chiyoda-ku, Tokyo 102-0082 (Japan)

2006-03-15

The value of the burnup is one of the most important parameters of samples taken by post-irradiation examination (PIE). Generally, it is evaluated by the Neodymium-148 method. Precise evaluation of the burnup value requires: (1) an effective fission yield of {sup 148}Nd; (2) neutron capture reactions of {sup 147}Nd and {sup 148}Nd; (3) a conversion factor from fissions per initial heavy metal to the burnup unit GWd/t. In this study, the burnup values of the PIE data from Mihama-3 and Genkai-1 PWRs, which were taken by the Japan Atomic Energy Research Institute, were re-evaluated using more accurate corrections for each of these three items. The PIE data were then re-analyzed using SWAT and SWAT2 code systems with JENDL-3.3 library. The re-evaluation of the effective fission yield of {sup 148}Nd has an effect of 1.5-2.0% on burnup values. Considering the neutron capture reactions of {sup 147}Nd and {sup 148}Nd removes dependence of C/E values of {sup 148}Nd on the burnup value. The conversion factor from FIMA(%) to GWd/t changes according to the burnup value. Its effect on the burnup evaluation is small for samples having burnup of larger than 30 GWd/t. The analyses using the corrected burnup values showed that the calculated {sup 148}Nd concentrations and the PIE data is approximately 1%, whereas this was 3-5% in prior analyses. This analysis indicates that the burnup values of samples from Mihama-3 and Genkai-1 PWRs should be corrected by 2-3%. The effect of re-evaluation of the burnup value on the neutron multiplication factor is an approximately 0.6% change in PIE samples having the burnup of larger than 30 GWd/t. Finally, comparison between calculation results using a single pin-cell model and an assembly model is carried out. Because the results agreed with each other within a few percent, we concluded that the single pin-cell model is suitable for the analysis of PIE samples and that the underestimation of plutonium isotopes, which occurred in the previous
Burn-up credit applications for UO2 and MOX fuel assemblies in AREVA/COGEMA

International Nuclear Information System (INIS)

Toubon, H.; Riffard, C.; Batifol, M.; Pelletier, S.

2003-01-01

For the last seven years, AREVA/COGEMA has been implementing the second phase of its burn-up credit program (the incorporation of fission products). Since the early nineties, major actinides have been taken into account in criticality analyses first for reprocessing applications, then for transport and storage of fuel assemblies Next year (2004) COGEMA will take into account the six main fission products (Rh103, Cs133, Nd143, Sm149, Sm152 and Gd155) that make up 50% of the anti-reactivity of all fission products. The experimental program will soon be finished. The new burn-up credit methodology is in progress. After a brief overview of BUC R and D program and COGEMA's application of the BUC, this paper will focus on the new burn-up measurement for UO2 and MOX fuel assemblies. It details the measurement instrumentation and the measurement experiments on MOX fuels performed at La Hague in January 2003. (author)

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)
Economic incentives and recommended development for commercial use of high burnup fuels in the once-through LWR fuel cycle

International Nuclear Information System (INIS)

Stout, R.B.; Merckx, K.R.; Holm, J.S.

1981-01-01

This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels
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
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)
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
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)
Advances in applications of burnup credit to enhance spent fuel transportation, storage, reprocessing and disposition. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2007-05-01

Given a trend towards higher burnup power reactor fuel, the IAEA began an active programme in burnup credit (BUC) with major meetings in 1997 (IAEA-TECDOC-1013), 2000 (IAEA-TECDOC-1241) and 2002 (IAEA-TECDOC-1378) exploring worldwide interest in using BUC in spent fuel management systems. This publication contains the proceedings of the IAEA's 4th major BUC meeting, held in London. Sixty participants from 18 countries addressed calculation methodology, validation and criticality, safety criteria, procedural compliance with safety criteria, benefits of BUC applications, and regulatory aspects in BUC. This meeting encouraged the IAEA to continue its activities on burnup credit including dissemination of related information, given the number of Member States having to deal with increased spent fuel quantities and extended durations. A 5th major meeting on burnup credit is planned 2008. Burnup credit is a concept that takes credit for the reduced reactivity of fuel discharged from the reactor to improve loading density of irradiated fuel assemblies in storage, transportation, and disposal applications, relative to the assumption of fresh fuel nuclide inventories in loading calculations. This report has described a general four phase approach to be considered in burnup credit implementation. Much if not all of the background research and data acquisition necessary for successful burnup credit development in preparation for licensing has been completed. Many fuel types, facilities, and analysis methods are encompassed in the public knowledge base, such that in many cases this guidance will provide a means for rapid development of a burnup credit program. For newer assembly designs, higher enrichment fuels, and more extensive nuclide credit, additional research and development may be necessary, but even this work can build on the foundation that has been established to date. Those, it is hoped that this report will serve as a starting point with sufficient reference to
Review of the effects of burnup on the thermal conductivity of UO2

International Nuclear Information System (INIS)

Lokken, R.O.; Courtright, E.L.

1976-01-01

The general trends which relate changes in thermal conductivity of UO 2 fuel as a function of temperature and burnup can be summarized as follows: (1) At temperatures below 500 0 C, reductions in UO 2 thermal conductivity relative to the unirradiated values can be expected up to a saturation level of approximately 10 19 fissions/cc. (2) At temperatures above 500 0 C, the thermal conductivity will undergo little change at low burnups, (less than 10 19 fissions/cc) but at higher exposures some decrease can be expected which should, in turn, diminish with increasing temperature. (3) A review of the data reported by Berman on the ThO 2 --UO 2 fuel indicates that the basic behavior is the same as for UO 2 in the temperature range of major interest. The applicability of this data to LWR UO 2 fuel is somewhat questionable because of basic physical property differences, and limited data on irradiation effects, and would not seem to support concerns that the effects of burnup on thermal conductivity for LWR fuel may be of more significance than currently believed. (4) A mathematical expression of the type proposed by Daniel and Cohen seems to provide a reasonable approximation for the behavioral trends reported in the literature which relate changes in thermal conductivity to increasing burnup in certain temperature regimes. Calculations indicate that only small incremental increases in the fuel centerline temperature might be expected if burnup effects are taken into account
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)
Comparative calculations and operation-to-PIE data juxtaposition of the Zaporozhye NPP, WWER-1000 FA-E0325 fuel rods after 4 years of operation up to ∼49 MWd/kgU burnup

International Nuclear Information System (INIS)

Passage, G.; Stefanova, S.; Scheglov, A.; Proselkov, V.

2006-01-01

Operational and PIE data for the Zaporozhe NPP, FA-E0325, WWER-1000 fuel rods were provided in the OECD NEA IFPE Database and were used to perform comparative calculations among several fuel performance codes. The fuel rods had been irradiated for 4 years of operation up to ∼49 MWd/kg U burnup. The fuel rod operation histories are developed for the PINw99, TRANSURANUS (V1M1J03) and TOPRA-2 codes. The initial state fuel rod parameters are analysed and calculations are carried out. The PIE data enable the comparison of experimental measurement with code-calculated values for cladding elongation (49 rods), FGR and gas pressure (35 rods). Cladding diameter creep-down and gap closure results are juxtaposed as well. The capability of the applied codes correctly to predict the WWER fuel rod performance is shown. The WWER-1000 fuel rod data include initial geometrical and design parameters of the fuel rods, as well as description of the operation regime, NPP unit loading history and PIE results at normal conditions. The data are sufficient for modelling all 312 fuel rod and for comparison of calculations with experimental results for a limited number of fuel rods. The comparison between the calculated and measured results discussed in this paper shows that the codes PINw99, TRANSURANUS and TOPRA-2, are capable of adequate predicting the thermophysical and the mechanical performance of the WWER-1000 fuel rods. The PINw99 code predicts conservative BOL FGR values and conservative gas pressure values in the region of burnups higher than 30 MWd/kg U, which can be explained by the underprediction of the cladding gas inner volume and cladding elongation. The improved version PIN2K (not applied in the present study) predicts much better FGR and gas pressure, though, it is still under development in the high burnup FGR modelling part. In the TRANSURANUS code, there are also areas, where refinements are clearly indicated. They are subjects of the ongoing research projects and
Reactivity and isotopic composition of spent PWR [pressurized-water-reactor] fuel as a function of initial enrichment, burnup, and cooling time

International Nuclear Information System (INIS)

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

1987-10-01

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

International Nuclear Information System (INIS)

Volkov, A.I.; Lukin, A.V.; Magda, L.E.; Magda, E.P.; Pogrebov, I.S.; Putnikov, I.S.; Khmelnitsky, D.V.; Scherbakov, A.P.

1998-01-01

A physical scheme of burnup reactor for a Periodic Pulsed Nuclear Pumped Laser was supposed. Calculations of its neutron physical parameters were made. The general layout and construction of basic elements of the reactor are discussed. The requirements for the fuel and fuel elements are established. (author)
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
Sensitivity change of rhodium self -powered detectors with burn-up

International Nuclear Information System (INIS)

Girgis, R.; Akimov, I.S.; Hamouda, I.

1976-01-01

The scope of the present paper is to obtain the calculation formulae to evaluate the rate of sensitivity change of the neutron self-powered detectors with burn-up. A code written in FORTRAN 4 was developed to be operational on the IBM-1130 computer. It has been established in the case of rhodium detectors that neglecting the β-particle absorption in the calculations leads to the underestimation of the detector sensitivity decrease up to 40%. The derived formulae can be used for other self-powered detectors. (author)
CONHOR. Code system for determination of power distribution and burnup for the HOR reactor. Version 1.0.. User`s manual

Energy Technology Data Exchange (ETDEWEB)

Serov, I V; Hoogenboom, J E

1993-07-01

The main calculational tool is the CITATION code. CITATION is used for both static and burnup calculations. The pointwise flux density and power distributions obtained from these calculations are used to obtain the values of the desired quantities at the beginning of a burnup cycle. To obtain the most trustful values of the desired quantities CONHOR employs experimental information together with the CITATION calculated flux distributions. Axially averaged foil activation rates are obtained based on both CITATION pointwise flux density distributions and measured foil activity counts. These two sets of activation rates are called the distributions of auxiliary quantities and are compared with each other in order to pick up the corrections to the U-235 number densities in fuel containing elements. The methodical corrections to the calculational auxiliary quantities are obtained on this basis as well. They are used to obtain the methodical corrections to the desired quantities. The corrected desired quantities are the recommended ones. The correction procedure requires the knowledge of the sensitivity coefficients of the average foil activation rates with respect to the U-235 number densities (through the text of this manual U-235 is denoted also and especially in the input-output description sections as a BUrning-COrrected material, or `BuCo` material). These sensitivity coefficients are calculated by the CONHOR SENS module. CITATION is employed to perform the calculations with perturbed values of U-235 number densities. Burnup calculations can be performed being based on either corrected or uncorrected U-235 number densities. Through the text of this manual XXXX means a 4-symbol identification of the burnup cycle to be studied. XX-1 and XX+1 mean correspondingly the previous and the following cycles. (orig./HP).
Application of depletion perturbation theory to fuel cycle burnup analysis

International Nuclear Information System (INIS)

White, J.R.

1979-01-01

Over the past several years static perturbation theory methods have been increasingly used for reactor analysis in lieu of more detailed and costly direct computations. Recently, perturbation methods incorporating time dependence have also received attention, and several authors have demonstrated their applicability to fuel burnup analysis. The objective of the work described here is to demonstrate that a time-dependent perturbation method can be easily and accurately applied to realistic depletion problems
Evaluation of burnup credit for accommodating PWR spent nuclear fuel in high-capacity cask designs

International Nuclear Information System (INIS)

Wagner, John C.

2003-01-01

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

The burn-up credit physics and the 40. Minerve anniversary

International Nuclear Information System (INIS)

Santamarina, A.; Toubon, H.; Trakas, C.

2000-01-01

The technical meeting organized by the SFEN on the burn-up credit (CBU) physics, took place the 23 november 1999 at Cadarache. the first presentation dealt with the economic interest and the neutronic problems of the CBU. Then two papers presented how taking into account the CBU in the industry in matter of transport, storage in pool, reprocessing and criticality calculation (MCNP4/Apollo2-F benchmark). An experimental method for the reactivity measurement through oscillations in the Minerve reactor, has been presented with an analysis of the possible errors. The future research program OSMOSE, taking into account the minor actinides in the CBU, was also developed. The last paper presented the national and international research programs in the CBU domain, in particular experiments realized in CEA/Valduc and the OECD Burn-up Criticality Benchmark Group activities. (A.L.B.)
Kinetic Monte Carlo Potts Model for Simulating a High Burnup Structure in UO2

International Nuclear Information System (INIS)

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

2008-01-01

A Potts model, based on the kinetic Monte Carlo method, was originally developed for magnetic domain evolutions, but it was also proposed as a model for a grain growth in polycrystals due to similarities between Potts domain structures and grain structures. It has modeled various microstructural phenomena such as grain growths, a recrystallization, a sintering, and so on. A high burnup structure (HBS) is observed in the periphery of a high burnup UO 2 fuel. Although its formation mechanism is not clearly understood yet, its characteristics are well recognized: The HBS microstructure consists of very small grains and large bubbles instead of original as-sintered grains. A threshold burnup for the HBS is observed at a local burnup 60-80 Gwd/tM, and the threshold temperature is 1000-1200 .deg. C. Concerning a energy stability, the HBS can be created if the system energy of the HBS is lower than that of the original structure in an irradiated UO 2 . In this paper, a Potts model was implemented for simulating the HBS by calculating system energies, and the simulation results were compared with the HBS characteristics mentioned above
Chemical analytical considerations on the determination of burnup in irradiated nuclear fuels

International Nuclear Information System (INIS)

Cretella, R.F.; Servant, R.E.

1989-01-01

Burnup in an irradiated nuclear fuel may be defined as the energy produced per mass unit, from the time the fuel is introduced into the reactor and until a given moment. It is usually shown in megawatt/day or megawatt/hour generated per ton or kilo of fuel. It is also indicated as the number of fission produced per volume unit (cm 3 ) or per every 100 initial fissionable atoms. The yield of a power plant is directly related to the burnup of its fuel load and knowing the latter contributes to optimizing the economy in reactor operation and the related technologies. The development of nuclear fuels and the operation of reactors require doing with exact and accurate methods allowing to know the burnup. Errors in this measurement have an incidence upon the fuel design, the physical and nuclear calculations, the shielding requirements, the design of vehicles for the transportation of irradiated fuels, the engineering of processing plants, etc. All these factors, in turn, have an incidence upon the cost of nuclear power generation. (Author) [es
Burnable poison calculations for Mk.III gas-cooled reactor systems

Energy Technology Data Exchange (ETDEWEB)

Gubbins, M E

1971-02-15

A method of calculating the reactivity and burn-up hisotry of a Mk.III GCR system containing burnable poisons has been described. The method allows for poison-fuel interaction. Using the method it has been shown that burn-up of the poison under a constant incident flux can give errors of the order of 1-2 niles. A calculation using the method described will take about 50% longer than a straightforward fuel burn-up calculation in the same number of groups. The multi-cell approach has a potential for handling greater geometrical complexity. It is intended to compare the method against experiment as soon as suitable experimental results become available.
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
Development of high performance liquid chromatography for rapid determination of burn-up of nuclear fuels

International Nuclear Information System (INIS)

Joseph, M.; Karunasagar, D.; Saha, B.

1996-01-01

Burn-up an important parameter during evaluation of the performance of any nuclear fuel. Among the various techniques available, the preferred one for its determination is based on accurate measurement of a suitable fission product monitor and the residual heavy elements. Since isotopes of rare earth elements are generally used as burn-up monitors, conditions were standardized for rapid separation (within 15 minutes) of light rare earths using high performance liquid chromatography based on either anion exchange (Partisil 10 SAX) in methanol-nitric acid medium or by cation exchange on a reverse phase column (Spherisorb 5-ODS-2 or Supelcosil LC-18) dynamically modified with 1-octane sulfonate or camphor-10-sulfonic acid (β). Both these methods were assessed for separation of individual fission product rare earths from their mixtures. A new approach has been examined in detail for rapid assay of neodymium, which appears promising for faster and accurate measurement of burn-up. (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)
MODRIB - a zero dimensional code for criticality and burn-up of LWR's

International Nuclear Information System (INIS)

Gaafar, M.A.; El-Cherif, A.I.

1980-01-01

The computer program MODRIB is a zero-dimensional code for calculating criticality and burn-up of light water reactors (LWR's). It is a version of an Italian code RIBOT-2 with an updated cross-section data library. The nuclear constants of MODRIB-code are calculated with a two group scheme (fast and thermal), where the fast group is an average of three fast groups. The code requires as input data essential extensive reactor parameters such as fuel rod radius, clad thickness, fuel enrichment, lattice pitch, water density and temperature etc. A summary of the physical model description and the input-output procedures are given in this report. Selected results of two sample problems are also given for the purpose of checking the validity and reliability of the code. The first is BWR and the second is PWR. The calculation time for a criticality problem with burn-up is about 8 seconds for the first time step and about 3 seconds for each subsequent time step on the ICL-1906 computer facility. The requirements on the memory size is less than 32 K-word. (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
Calculation using MVP and MVP-BURN in JRR-3

International Nuclear Information System (INIS)

Komeda, Masao; Kato, Tomoaki; Murayama, Yoji; Yamashita, Kiyonobu

2007-01-01

MVP is the particle-transport Monte Carlo code that has been developed in JAEA. MVP-BURN is an added function to do burn-up calculation. It is easy to built complex structure like core for MVP. And it is easy to do calculations of keff, any reaction rate, flux, burn-up and so on. In this report, it is introduced MVP and MVP-BURN. And some sample calculations of JRR-3 are shown. (author)
Modelling of high burnup structure in UO2 fuel with the RTOP code

International Nuclear Information System (INIS)

Likhanskii, V.; Zborovskii, V.; Evdokimov, I.; Kanyukova, V.; Sorokin, A.

2008-01-01

The present work deals with self-consistent physical approach aimed to derive the criterion of fuel restructuring avoiding correlations. The approach is based on study of large over pressurized bubbles formation on dislocations, at grain boundaries and in grain volume. At first, stage of formation of bubbles non-destroyable by fission fragments is examined using consistent modelling of point defects and fission gas behavior near dislocation and in grain volume. Then, evolution of formed large non-destroyable bubbles is considered using results of the previous step as initial values. Finally, condition of dislocation loops punching by sufficiently large over pressurized bubbles is regarded as the criterion of fuel restructuring onset. In the present work consideration of large over pressurized bubbles evolution is applied to modelling of the restructuring threshold depending on temperature, burnup and grain size. Effect of grain size predicted by the model is in qualitative agreement with experimental observations. Restructuring threshold criterion as an analytical function of local burnup and fuel temperature is derived and compared with HBRP project data. To predict rim-layer width formation depending on fuel burnup and irradiation conditions the model is implemented into the mechanistic fuel performance code RTOP. Calculated dependencies give upper estimate for the width of restructured region. Calculations show that one needs to consider temperature distribution within pellet which depends on irradiation history in order to model rim-structure formation
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)
Evaluation and reffinement of the neutronic calculation methodology

International Nuclear Information System (INIS)

Conti Filho, P.

1984-01-01

A computational code that has the homogenized cross section given by the LEOPARD code as input was developed. The code gives polinomial coefficients that represent the homogenized cross section as a function of the local burnup and the boron concentration for the assembly, for each step in the reactor Burnup. Lately, were developed an interface between the LEOPARD code Polinomiun Generator program and CITATION code to became possible to CITATION code to set the homogenized microscopic cross section as function of the local caracteristics of the assembly on the way to make the calculation of the reactor Burnup. For a choosen reactor (1900MWth) have been done the inicial calculation (super-cells calculation and others Input) and after that were done the calculation with and without the polinomia. The analyses of the results of the CITATION code were done and the principal results were presented here. (Author) [pt
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
Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

International Nuclear Information System (INIS)

Ozdemir, Levent; Acar, Banu Bulut; Zabunoglu, Okan H.

2011-01-01

When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of 239 Pu and 241 Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.
MCOR - Monte Carlo depletion code for reference LWR calculations

Energy Technology Data Exchange (ETDEWEB)

Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)

2011-04-15

Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally
MCOR - Monte Carlo depletion code for reference LWR calculations

International Nuclear Information System (INIS)

Puente Espel, Federico; Tippayakul, Chanatip; Ivanov, Kostadin; Misu, Stefan

2011-01-01

Research highlights: → Introduction of a reference Monte Carlo based depletion code with extended capabilities. → Verification and validation results for MCOR. → Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

International Nuclear Information System (INIS)

Cepraga, D.; Boeriu, St.; Gheorghiu, E.; Cristian, I.; Patrulescu, I.; Cimporescu, D.; Ciuvica, P.; Velciu, E.

1975-01-01

The calculation system DACC-5 is a zero-dimensional reactor physics code used to calculate the criticality and burn-up of light-water reactors. The code requires as input essential extensive reactor parameters (fuel rod radius, water density, etc.). The nuclear constants (intensive parameters) are calculated with a five-group model (2 thermal and 3 fast groups). A fitting procedure is systematically employed to reduce the computation time of the code. Zero-dimensional burn-up calculations are made in an automatic way. Part one of the paper contains the code physical model and computer structure. Part two of the paper will contain tests of DACC-5 credibility for different light-water power lattices
A comparison study of the 1MeV triton burn-up in JET using the HECTOR and SOCRATE codes

International Nuclear Information System (INIS)

Gorini, G.; Kovanen, M.A.

1988-01-01

The burn-up of the 1MeV tritons in deuterium plasmas has been measured in JET for various plasma conditions. To interpret these measurements the containment, slowing down and burn-up of fast tritons needs to be modelled with a reasonable accuracy. The numerical code SOCRATE has been written for this specific purpose and a second code, HECTOR, has been adapted to study the triton burn-up problem. In this paper we compare the results from the two codes in order to exclude possible errors in the numerical models, to assess their accuracy and to study the sensitivity of the calculation to various physical effects. (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 performance of OTTO cycle pebble bed reactors with ROX fuel

International Nuclear Information System (INIS)

Ho, Hai Quan; Obara, Toru

2015-01-01

Highlights: • A 300 MW t Small Pebble Bed Reactor with Rock-like oxide fuel is proposed. • Using ROX fuel can achieve high discharged burnup of spent fuel. • High geological stability can be expected in direct disposal of the spent ROX fuel. • The Pebble Bed Reactor with ROX fuel can be critical at steady state operation. • All the reactor designs have a negative temperature coefficient. - Abstract: A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO 2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO 2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO 2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor
Effect of Pu-rich agglomerate in MOX fuel on a lattice calculation

International Nuclear Information System (INIS)

Kawashima, Katsuyuki; Yamamoto, Toru; Namekawa, Masakazu

2007-01-01

The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) fuel on a lattice calculation has been demonstrated. The Pu-rich agglomerate parameters are defined based on the measurement data of MIMAS-MOX and the focus is on the highly enriched MOX fuel in accordance with increased burnup resulting in a higher volume fraction of the Pu-rich agglomerates. The lattice calculations with a heterogeneous fuel model and a homogeneous fuel model are performed simulating the PWR 17x17 fuel assembly. The heterogeneous model individually treats the Pu-rich agglomerate and U-Pu matrix, whereas the homogeneous model homogenizes the compositions within the fuel pellet. A continuous-energy Monte Carlo burnup code, MVP-BURN, is used for burnup calculations up to 70 GWd/t. A statistical geometry model is applied in modeling a large number of Pu-rich agglomerates assuming that they are distributed randomly within the MOX fuel pellet. The calculated nuclear characteristics include k-inf, Pu isotopic compositions, power density and burnup of the Pu-rich agglomerates, as well as the pellet-averaged Pu compositions as a function of burnup. It is shown that the effect of Pu-rich agglomerates on the lattice calculation is negligibly small. (author)
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
Study of the influence of slab perturbation in the cell on the fuel local burnup

International Nuclear Information System (INIS)

Takac, S.; Kocic, A.; Dimitrijevic, Z.; Markovic, H.; Dimitrijevic, V.

1975-01-01

The influence of construction material or voids in the fuel element on the fuel burnup was the objective of this study. Experiments were done by cell perturbation method. Theoretical method was developed for calculating the effect of reactor cell perturbation. Obtained results both experimental and theoretical clearly indicate that the minimum quantity of construction material or void cause local increase of neutron flux in the mentioned regions. This increase of flux which amounts to nearly ten percent, and can reach the value of a few tens percent leads to the local increase of fuel burnup [sr
3-D full core calculations for the long-term behaviour of PWR's

International Nuclear Information System (INIS)

Winter, H.J.; Koebke, K.; Wagner, M.R.

1987-01-01

Presently, the most realistic simulation of a PWR core is by means of three-dimensional (3-D) full core calculations. Only by such 3-D representations can the large scope of axial effects be treated in an accurate and direct way, without the need to perform various auxiliary calculations. Although the computationally efficient burnup-corrected nodal expansion method (NEM-BC) is used, the computing effort for 3-D reactor calculations becomes rather high, e.g. a storage of about 320000 words is required to describe a 1300 MWe PWR. NEM-BC was introduced (1979) into KWU's package of PWR design codes because of its high accuracy and the great reduction of computing time and storage requirements in comparison to other methods. The application of NEM-BC to 3-dimensional PWR design is strongly correlated with the progress achieved in the solution of the homogenization and dehomogenization problem. By means of suitable methods (equivalence theory) the transport-theoretical information of the pinwise power and burnup distribution for the heterogeneous fuel assemblies is transferred in a consistent manner to the full core reactor solution. The new methods and the corresponding code system are explained in some detail. (orig.)
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
Preliminary assessment of the benefits of derating a cask for increasing age/burnup capability

International Nuclear Information System (INIS)

Broadhead, B.L.; Parks, C.V.; Joy, D.S.; Tang, J.S.

1992-01-01

This paper discusses a study performed to determine the extent to which the age/burnup capability of the Babcock and Wilcox BR-100 rail cask could be extended by reducing the number of fuel assemblies. Only the shielding effects of derating are accounted for in this study. Separate analyses will be necessary to address the enhanced heat loads due to increased burnup or decreased age. The criterion used to assess the derating was the calculated dose 2 m from the rail car. The reference calculations were based on the 70% design of the BR-100 cask with 21 PWR fuel assemblies. Seven different basket/assembly loading configurations were investigated. The results indicate that both an alternative 18-assembly basket configuration and a 17-assembly/4-empty-hole configuration for the 21-element basket offer substantial gains over the fully loaded reference 21-element basket configuration
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
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)
A validated methodology for evaluating burn-up credit in spent fuel casks

International Nuclear Information System (INIS)

Brady, M.C.; Sanders, T.L.

1992-01-01

The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (USDOE) programme to resolve issues related to the implementation of burn-up credit in spent fuel cask design. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burn-up credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor re-start critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias effective multiplication (k eff ). Implementation issues affecting licensing requirements and operational procedures are discussed briefly. (Author)
Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

Directory of Open Access Journals (Sweden)

Asah-Opoku Fiifi

2014-01-01

Full Text Available Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX, uranium oxide fuel (UOX, and commercially enriched uranium or uranium metal (CEU - are used in this simulation and their impact on the effective multiplication factor (Keff and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.
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.)
Determination of depth of burnup of fuel in deposition in territory Belarus, after Chernobyl accident, with the help of a tracer of U-236

International Nuclear Information System (INIS)

Mironov, V.P.; Matusevich, Zh.L.; Kudryashov, V.P.; Ananich, P.I.; Zhuravkov, V.V.

2002-01-01

Experiments and calculations for determination of depth of burnup of fuel are carried out on separate sites in Belarus. As a tracer of Chernobyl deposition the uranium-236 was used. The average depth of burnup of fuel in 30 km zone is 9,4 MW*d/kgU

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)
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.)
EPRI/DOE High Burnup Fuel Sister Pin 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, Brady [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)

2017-07-01

The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.
Fuel burn-up distribution and transuranic nuclide contents produced at the first cycle operation of AP1000

International Nuclear Information System (INIS)

Jati Susilo; Jupiter Sitorus Pane

2016-01-01

AP1000 reactor core was designed with nominal power of 1154 MWe (3415 MWth), operated within life time of 60 years and cycle length of 18 months. For the first cycle, the AP1000 core uses three kinds of UO 2 enrichment, they are 2.35 w/o, 3.40 w/o and 4.45 w/o. Absorber materials such as ZrB 2 , Pyrex and Boron solution are used to compensate the excess reactivity at the beginning of cycle. In the core, U-235 fuels are burned by fission reaction and produce energy, fission products and new neutron. Because of the U-238 neutron absorption reaction, the high level radioactive waste of heavy nuclide transuranic such as Pu, Am, Cm and Np are also generated. They have a very long half life. The purpose of this study is to evaluate the result of fuel burn-up distribution and heavy nuclide transuranic contents produced by AP1000 at the end of first cycle operation (EOFC). Calculation of ¼ part of the AP1000 core in the 2 dimensional model has been done using SRAC2006 code with the module of COREBN/HIST. The input data called the table of macroscopic cross section, is calculated using module of PIJ. The result shows that the maximum fuel assembly (FA) burn-up is 27.04 GWD/MTU, that is still lower than allowed maximum burn-up of 62 GWD/MTU. Fuel loading position at the center/middle of the core will produce bigger burn-up and transuranic nuclide than one at the edges the of the core. The use of IFBA fuel just give a small effect to lessen the fuel burn-up and transuranic nuclide production. (author)
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
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
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.)
Investigation of research and development subjects for the Very High Burnup Fuel

International Nuclear Information System (INIS)

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

1993-06-01

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

International Nuclear Information System (INIS)

Hill, R.N.; Kawashima, M.; Arie, K.; Suzuki, M.

1992-01-01

Recently, a number of low void worth core designs with non-conventional core geometries have been proposed. Since these designs lack a good experimental and computational database, benchmark calculations are useful for the identification of possible biases in performance characteristics predictions. In this paper, a simplified benchmark model of a metal fueled, low void worth actinide burner design is detailed; and two independent neutronic performance evaluations are compared. Calculated performance characteristics are evaluated for three spatially uniform compositions (fresh uranium/plutonium, batch-averaged uranium/transuranic, and batch-averaged uranium/transuranic with fission products) and a regional depleted distribution obtained from a benchmark depletion calculation. For each core composition, the flooded and voided multiplication factor, power peaking factor, sodium void worth (and its components), flooded Doppler coefficient and control rod worth predictions are compared. In addition, the burnup swing, average discharge burnup, peak linear power, and fresh fuel enrichment are calculated for the depletion case. In general, remarkably good agreement is observed between the evaluations. The most significant difference is predicted performance characteristics is a 0.3--0.5% Δk/(kk) bias in the sodium void worth. Significant differences in the transmutation rate of higher actinides are also observed; however, these differences do not cause discrepancies in the performing predictions
3-DB, 3-D Multigroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup

International Nuclear Information System (INIS)

Hardie, R.W.; Little, W.W. Jr.; Mroz, W.

1974-01-01

1 - Description of problem or function: 3DB is a three-dimensional (x-y-z, r-theta-z, triangular-z) multigroup diffusion code for use in detailed fast-reactor criticality and burnup analysis. The code can be used to - (a) compute k eff and perform criticality searches on time absorption, reactor composition, and reactor dimensions by means of either a flux or an adjoint model, (b) compute material burnup using a flexible material shuffling scheme, and (c) compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Eigenvalues are computed by standard source- iteration techniques. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Adjoint solutions are obtained by inverting the input data and redefining the source terms. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy-averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes are formed by the user. The code does not contain built- in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated
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)
Portable gamma-ray holdup and attributes measurements of high- and variable-burnup plutonium

International Nuclear Information System (INIS)

Wenz, T.R.; Russo, P.A.; Miller, M.C.; Menlove, H.O.; Takahashi, S.; Yamamoto, Y.; Aoki, I.

1991-01-01

High burnup-plutonium holdup has been assayed quantitatively by low resolution gamma-ray spectrometry. The assay was calibrated with four plutonium standards representing a range of fuel burnup and 241 Am content. Selection of a calibration standard based on its qualitative spectral similarity to gamma-ray spectra of the process material is partially responsible for the success of these holdup measurements. The spectral analysis method is based on the determination of net counts in a single spectral region of interest (ROI). However, the low-resolution gamma-ray assay signal for the high-burnup plutonium includes unknown amounts of contamination from 241 Am. For most needs, the range of calibration standards required for this selection procedure is not available. A new low-resolution gamma-ray spectral analysis procedure for assay of 239 Pu has been developed. The procedure uses the calculated isotope activity ratios and the measured net counts in three spectral ROIs to evaluate and remove the 241 Am contamination from the 239 Pu assay signal on a spectrum-by-spectrum basis. The calibration for the new procedure requires only a single plutonium standard. The procedure also provides a measure of the burnup and age attributes of holdup deposits. The new procedure has been demonstrated using portable gamma-ray spectroscopy equipment for a wide range of plutonium standards and has also been applied to the assay of 239 Pu holdup in a mixed oxide fuel fabrication facility. 10 refs., 5 figs., 3 tabs
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
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
Numeric modeling of HfO2 neutron flux sensor parameters during sensor burnup in the RBMK-1500 reactor

International Nuclear Information System (INIS)

Jurkevicius, A.; Remeikis, V.

2001-01-01

The isotopic composition of hafnium in the radial neutron flux sensor of the RBMK-1500 reactor, the rates of the neutron absorption on Hf isotopes and the neutron spectrum in the sensor were numerically modeled. The sequence SAS2 (Shielding Analysis Sequence) from the package SCALE 4.3 was used for calculations. It has been obtained that the main neutron absorber 167 Er isotope practically burns up completely at the 18 MW d/kgU burnup depth, and at that time the capture rate of thermal neutrons in erbium decreases ten-fold. The average neutron flux density was calculated 7.6*10 13 neutrons. Cm -2 S -1 in the RBMK-1500 reactor grating, when the nuclear fuel enriched with 235 U by 2.4% and with Er by 0.4% is used in a fuel assembly. When the sensor burnup reaches 28 MW d/kgU, the neutron absorption rate of 178 Hf exceeds the rate of 177 Hf. The overall neutron absorption rate in hafnium decreases 2.53 times due to the sensor burnup to 56 MW d/kgU. The corrective factors ξ d (I) at different integral flux I of the sensors were calculated. The obtained dependence ξ d (I) calculated numerically was compared to the experimental one determined by processing repeated calibration results of Hf sensors in RBMK-1500 reactors, as well as compared to the theoretical one currently used in the Ignalina NPP special mathematical algorithms. (author)
Some Thermodynamic Features of Uranium-Plutonium Nitride Fuel in the Course of Burnup

Science.gov (United States)

Rusinkevich, A. A.; Ivanov, A. S.; Belov, G. V.; Skupov, M. V.

2017-12-01

Calculation studies on the effect of carbon and oxygen impurities on the chemical and phase compositions of nitride uranium-plutonium fuel in the course of burnup are performed using the IVTANTHERMO code. It is shown that the number of moles of UN decreases with increasing burnup level, whereas UN1.466, UN1.54, and UN1.73 exhibit a considerable increase. The presence of oxygen and carbon impurities causes an increase in the content of the UN1.466, UN1.54 and UN1.73 phases in the initial fuel by several orders of magnitude, in particular, at a relatively low temperature. At the same time, the presence of impurities abruptly reduces the content of free uranium in unburned fuel. Plutonium in the considered system is contained in form of Pu, PuC, PuC2, Pu2C3, and PuN. Plutonium carbides, as well as uranium carbides, are formed in small amounts. Most of the plutonium remains in the form of nitride PuN, whereas unbound Pu is present only in the areas with a low burnup level and high temperatures.
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

Extension of the Th-232 burnup chain in the WIMSD/4 program library

International Nuclear Information System (INIS)

Caldeira, A.D.

1991-07-01

The Th-232 burnup chain was extended through U-236, in the WIMSD/4 program library. The evolution of the values of k i nf and U-235 number density, as function of time, for the modified TRX1 problem, calculated with the new library, shows an improvement in the results when compared with LEOPARD program. (author)
Pu-rich MOX agglomerate-by-agglomerate model for fuel pellet burnup analysis

International Nuclear Information System (INIS)

Chang, G.S.

2004-01-01

In support of potential licensing of the mixed oxide (MOX) fuel made from weapons-grade (WG) plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The WG-MOX comprises five percent PuO 2 and 95% depleted UO 2 . Based on the Post Irradiation Examination (PIE) observation, the volume fraction (VF) of MOX agglomerates in the fuel pellet is about 16.67%, and PuO 2 concentration of 30.0 = (5 / 16.67 x 100) wt% in the agglomerate. A pressurized water reactor (PWR) unit WG-MOX lattice with Agglomerate-by-Agglomerate Fuel (AbAF) modeling has been developed. The effect of the irregular agglomerate distribution can be addressed through the use of the Monte Carlo AbAF model. The AbAF-calculated cumulative ratio of Agglomerate burnup to U-MAtrix burnup (AG/MA) is 9.17 at the beginning of life, and decreases to 2.88 at 50 GWd/t. The MCNP-AbAF-calculated results can be used to adjust the parameters in the MOX fuel fission gas release modeling. (author)
Experimental programmes related to high burnup fuel

International Nuclear Information System (INIS)

Vasudeva Rao, P.R.; Vidhya, R.; Ananthasivan, K.; Srinivasan, T.G.; Nagarajan, K.

2002-01-01

The experimental programmes undertaken at IGCAR with regard to high burn-up fuels fall under the following categories: a) studies on fuel behaviour, b) development of extractants for aqueous reprocessing and c) development of non-aqueous reprocessing techniques. An experimental programme to measure the carbon potential in U/Pu-FP-C systems by methane-hydrogen gas equilibration technique has been initiated at IGCAR in order to understand the evolution of fuel and fission product phases in carbide fuel at high burn-up. The carbon potentials in U-Mo-C system have been measured by this technique. The free energies and enthalpies of formation of LaC 2 , NdC 2 and SmC 2 have been measured by measuring the vapor pressures of CO over the region Ln 2 O 3 -LnC 2 -C during the carbothermic reduction of Ln 2 O 3 by C. The decontamination from fission products achieved in fuel reprocessing depends strongly on the actinide loading of the extractant phase. Tri-n-butyl phosphate (TBP), presently used as the extractant, does not allow high loadings due to its propensity for third phase formation in the extraction of Pu(IV). A detailed study of the allowable Pu loadings in TBP and other extractants has been undertaken in IGCAR, the results of which are presented in this paper. The paper also describes the status of our programme to develop a non-aqueous route for the reprocessing of fast reactor fuels. (author)
Determination of burnup in irradiated nuclear fuels by the method of stable 148Nd fission products

International Nuclear Information System (INIS)

Souza Sarkis, J.E. de.

1982-01-01

A method is described for the isotopic analysis and determination of burnup in irradiated nuclear fuel by mass spectrometric technique. The burnup is calculed from the determination of the concentration of uranium, plutonium and the fission product 148 Nd in the samples of UO 2 irradiated fuel from a P.W.R. type reactor. The method involves the separation of fractions of uranium, plutonium and neodimium from the dissolved irradiated fuel by ion exchange technique. The determination of uranium, plutonium and the fission product 148 Nd is carried out by isotope diluition mass spectrometry technique using as isotope tracers the nuclides 233 U, 242 Pu and 150 Nd. For the chemical processing and handling of the irradiated sample a Hot Chemistry Laboratory was mounted. It consists of glove boxes and equipments for radiation monitoring and protection. The results obtained indicate an atom percent burnup of 2.181 + - 0.035% wich corresponds an 20.937 + - 0.739 Gwd/ton of thermal energy produced. (Author) [pt
Evaluation of burnup credit for fuel storage analysis -- Experience in Spain

International Nuclear Information System (INIS)

Conde, J.M.; Recio, M.

1995-01-01

Several Spanish light water reactor commercial nuclear power plants are close to maximum spent-fuel pool storage capacity. The utilities are working on the implementation of state-of-the-art methods to increase the storage capacity, including both changes in the pool design (recracking) and the implementation of new analysis approaches with reduced conservation (burnup credit). Burnup credit criticality safety analyses have been approved for two pressurized water reactor plants (four units) and one boiling water reactor (BWR); an other BWR storage analysis is being developed at this moment. The elimination of the ''fresh fuel assumption'' increases the complexity of the criticality analysis to be performed, sometimes putting into question the capability of the analytic tools to properly describe this new situation and increasing the scope of the scenarios to be analyzed. From a regulatory perspective, the reactivity reduction associated with burnup of the fuel can be given credit only if the exposure of each fuel bundle can be known with enough accuracy. Subcriticality of spent-fuel storage depends mainly on the initial fuel enrichment, storage geometry, fuel exposure history, and cooling time. The last two aspects introduced new uncertainties in the criticality analysis that should be quantified in an adequate way. In addition, each and every fuel bundle has its own specific exposure history, so that strong assumptions and simplified calculational schemes have to be developed to undertake the analysis. The Consejo de Seguridad Nuclear (CSN), Spanish regulatory authority on the matter of nuclear safety and radiation protection, plays an active role in the development of analysis methods to support burnup credit, making proposals that may be beneficial in terms of risk and cost while keeping the widest safety margins possible
Analytical calculation of detailed model parameters of cast resin dry-type transformers

International Nuclear Information System (INIS)

Eslamian, M.; Vahidi, B.; Hosseinian, S.H.

2011-01-01

Highlights: → In this paper high frequency behavior of cast resin dry-type transformers was simulated. → Parameters of detailed model were calculated using analytical method and compared with FEM results. → A lab transformer was constructed in order to compare theoretical and experimental results. -- Abstract: Non-flammable characteristic of cast resin dry-type transformers make them suitable for different kind of usages. This paper presents an analytical method of how to obtain parameters of detailed model of these transformers. The calculated parameters are compared and verified with the corresponding FEM results and if it was necessary, correction factors are introduced for modification of the analytical solutions. Transient voltages under full and chopped test impulses are calculated using the obtained detailed model. In order to validate the model, a setup was constructed for testing on high-voltage winding of cast resin dry-type transformer. The simulation results were compared with the experimental data measured from FRA and impulse tests.
Verification to the RSG-GAS fuel discharge burn-up using SRAC2006 module of COREBN/HIST

International Nuclear Information System (INIS)

J-Susilo; T-M-Sembiring; G-R-Sunaryo; M-Imron

2018-01-01

For 30 years operation, some of the modifications to the RSG GAS core has been done, that are changes included the type of fuel from U 3 O 8 -Al to U 3 Si 2 -Al with the same density 2.96 gU/cc, the loading pattern of standard fuel elements/fuel control elements from 6/1 & 6/2 to 5/1 pattern, and in core fuel management calculation tool has been change from IAFUEL to BATAN-FUEL. To obtain an extension of the operating license for the next 10 years, the RSG-GAS Periodic Safety Assessment Document is need to prepared. According to the Regulatory Body Chairman Regulation No. 2 2015, RSG-GAS safety assessment should be done independently. As part of this assessment the fuel discharge burn-up must be estimated. In this research, to ensure that the misposition of fuel element in the core has not occurred, the investigation to the document operating report related the fuel placement has been done. Therefore, by using 78 th to 93 rd operation data, verify of the fuel discharge burn-up of the RSG-GAS has been performed by using SRAC2006 module of COREBN/HIST. In addition, the results of these calculations are also made comparative with the operating report data that is calculated by using BATAN-FUEL. Maximum fuel discharge burn-up (57.73 % of U-235) was verified still under permissible value determined by the regulatory body (<60 % of U-235). Maximum differences value between two computer codes was about 2.12 % of U-235 (3.80 %) that is fuel at the B-7 position. Fuel discharge burn-up of RSG-GAS showed almost the same value for each the operation cycle, range of 1.52 % of U-235. So it can be concluded that the RSG-GAS core operation over the last ten years was in good fuel management performance, in accordance with the design. BATAN-FUEL has been conformed well enough with COREBN/HIST. (author)
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
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
Impact on burnup performance of coated particle fuel design in pebble bed reactor with ROX fuel

International Nuclear Information System (INIS)

Ho, Hai Quan; Obara, Toru

2015-01-01

The pebble bed reactor (PBR), a kind of high-temperature gas-cooled reactor (HTGR), is expected to be among the next generation of nuclear reactors as it has excellent passive safety features, as well as online refueling and high thermal efficiency. Rock-like oxide (ROX) fuel has been studied at the Japan Atomic Energy Agency (JAEA) as a new once-through type fuel concept. Rock-like oxide used as fuel in a PBR can be expected to achieve high burnup and improve chemical stabilities. In the once-through fuel concept, the main challenge is to achieve as high a burnup as possible without failure of the spent fuel. The purpose of this study was to investigate the impact on burnup performance of different coated fuel particle (CFP) designs in a PBR with ROX fuel. In the study, the AGR-1 Coated Particle design and Deep-Burn Coated Particle design were used to make the burnup performance comparison. Criticality and core burnup calculations were performed by MCPBR code using the JENDL-4.0 library. Results at equilibrium showed that the two reactors utilizing AGR-1 Coated Particle and Deep-Burn Coated Particle designs could be critical with almost the same multiplication factor k eff . However, the power peaking factor and maximum power per fuel ball in the AGR-1 coated particle design was lower than that of Deep-Burn coated particle design. The AGR-1 design also showed an advantage in fissions per initial fissile atoms (FIFA); the AGR-1 coated particle design produced a higher FIFA than the Deep-Burn coated particle design. These results suggest that the difference in coated particle fuel design can have an effect on the burnup performance in ROX fuel. (author)
Nuclear criticality safety: general. 5. Reactivity Effect of Burnable Absorbers in Burnup Credit for the CASTOR X/32S Storage and Transport Cask

International Nuclear Information System (INIS)

Rombough, Charles T.; Lancaster, Dale B.; Diersch, Rudolf; Spilker, Harry

2001-01-01

When considering burnup credit in the licensing of storage and transportation casks, a significant effect is whether or not the burned fuel was depleted with burnable absorbers present. This paper presents the results of detailed calculations to quantitatively determine the burnable absorber effect for the CASTOR X/32S transport cask, which assumes burnup of the fuel in the criticality analysis. A radial view of the CASTOR X/32S cask is shown in Fig. 1. This is the actual plot of the geometry as modeled in KENO V.a. Note that there are no water-filled flux traps and the assemblies are tightly packed. This reduces the overall dimensions of the cask for a given number of fuel assemblies. Reactivity is held down by borated aluminum plates between the fuel assemblies and by placing absorber rod modules (ARMs) in the guide tubes of selected assemblies. If burnup of the fuel is not considered and the initial enrichment is 5.0 wt% 235 U, then 28 of the 32 fuel assemblies must contain an ARM to maintain a k eff 3 ; 4. moderator temperature of 604 K; 5. cooling time of 9.5 yr; 6. specific power of 60 W/g of U metal; 7. conservative axial and radial burnup shape distribution; 8. Westinghouse BP material containing 12.5 wt% B 4 C. Using the model described earlier, calculations were performed with varying numbers of BP fingers inserted for different exposure times. The results are shown in Tables I and II. The 1 s statistical error in these results is σ equals ±0.05%. Note that the BP finger and exposure effects decrease with fuel burnup and the effect is smaller when the cask contains ARMs. Conservatively combining the results from Tables I and II and interpolating, we can equate fewer BP fingers with longer BP exposure time as shown in Table III. The Table III results were checked by running the actual cases (for example, 20 BP fingers for 24 GWd/tonne exposure) to verify that the k eff 's for the cask were always less than the base-case values. These results can also be
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)
The role of ORIGEN-S in the design of burnup credit spent fuel casks

International Nuclear Information System (INIS)

Brady, M.C.

1991-01-01

Current licensing practices for spent fuel pools, storage facilities, and transportation casks require a conservative ''fresh fuel assumption'' be used in the criticality analysis. Burnup credit refers to a new approach in criticality analyses for spent fuel handling systems in which reactivity credit is allowed for the depleted state of the fuel. Studies have shown that the increased cask capacities that can be achieved with burnup credit offer both economic and risk incentives. The US Department of Energy is currently sponsoring a program to develop analysis methodologies and establish a new generation of spent fuel casks using the principle of burnup credit. The key difference in this new approach is the necessity to accurately predict the isotopic composition of the spent fuel. ORIGEN-S was selected to satisfy this requirement because of the flexibility and user-friendly input offered via its usage in the Standardized Computer Analyses for Licensing and Evaluation (SCALE) code system. Specifically, through the Shielding Analysis Sequence 2H (SAS2H), ORIGEN-S is linked with cross-section processing codes and one-dimensional transport analyses to produce problem-specific cross-section data for the point-depletion calculation. The utility code COUPLE facilitates updating basic cross-section and fission-yield data for the calculations. This paper describes the fundamental role fulfilled by ORIGEN-S in the development of the analysis methodology, validation of the methods, definition of criticality safety margins and other licensing considerations in the design of a new generation of spent fuel casks. Particular emphasis is given to the performance of ORIGEN-S in comparisons with measurements of irradiated fuel compositions and in predicting isotopics for use in the calculation of reactor restart critical configurations that are performed as a part of the validation process
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.
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
Nondestructive analysis of RA reactor fuel burnup, Program for burnup calculation base on relative yield of {sup 106}Ru, {sup 134}Cs and {sup 137}Cs in the irradiated fuel; Nedestruktivno odredjivanje izgaranja goriva reaktora RA, Program za izracunavanje izgaranja na osnovu relativne zastupljenosti {sup 106}Ru, {sup 134}Cs i {sup 137}Cs u ozracenom gorivu

Energy Technology Data Exchange (ETDEWEB)

Bulovic, V F [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1971-07-01

Burnup of low enriched metal uranium fuel of the RA reactor is described by two chain reactions. Energy balance and material changes in the fuel are described by systems of differential equations. Numerical integration of these equations is base on the the reactor operation data. Neutron flux and percent of Uranium-235 or more frequently yield of epithermal neutrons in the neutron flux, is determined by iteration from the measured contents of {sup 106}Ru, {sup 134}Cs and {sup 137}Cs in the irradiated fuel. The computer program was written in FORTRAN-IV. Burnup is calculated by using the measured activities of fission products. Burnup results are absolute values. Sagorevanje maloobogacenog uranskog metalnog goriva reaktora RA je opisano dvema lancanim reakcijama. Energetski bilans i materijalne promene u gorivu su opisane sistemima diferencijalnih jednacina. Numericka integracija jednacina se vrsi na osnovu podataka u dinamici rada reaktora. Fluks reaktorskih neutrona i procenat urana-235 ili ucesce epitermalnih neutrona u fluksu, odredjuje se iterativno na osnovu izmerenog sadrzaja {sup 106}Ru, {sup 134}Cs i {sup 137}Cs u ozracenom gorivu. Program je napisan u FORTRAN-u IV u jednom bloku, bez podprograma. Izracunavanje izgaranja je zasnovano na izmerenim kolicnicima aktivnosti fisionih produkata. Rezultati izgaranja imaju apsolutni karakter (author)
Nuclear calculation methods for light water moderated reactors

International Nuclear Information System (INIS)

Hicks, D.

1961-02-01

This report is intended as an introductory review. After a brief discussion of problems encountered in the nuclear design of water moderated reactors a comprehensive scheme of calculations is described. This scheme is based largely on theoretical methods and computer codes developed in the U.S.A. but some previously unreported developments made in this country are also described. It is shown that the effective reproduction factor of simple water moderated lattices may be estimated to an accuracy of approximately 1%. Methods for treating water gap flux peaking and control absorbers are presented in some detail, together with a brief discussion of temperature coefficients, void coefficients and burn-up problems. (author)
Investigation of the burn-up behavior of boron poison rods, placed in a fuel assembly of a pressurized water reactor

International Nuclear Information System (INIS)

Arnold, C.; Lutz, D.C.

1979-09-01

The excess reactivity of a pressurized water reactor is compensated by boron, disolved in the moderator. In addition during the first cycle boron poison rods are placed in fuel assemblies without control rods. The burn-up behavior of a poison rod in a Biblis B fuel assembly is analysed in the present paper. Multigroup spectrum calculations were performed. The influence of critical boron concentration depending from burn-up, the changes of fuel concentration and the concentration of burnable poison were taken into consideration. Furthermore the built-up of rapidly saturating fisson products 135 Xe and 149 Sm was considered. The interaction of these effects are discussed. Spatial influences are emphasized most. Finally two group cross sections were calculated. The results are compared with calculations for a fuel assembly of the same type without burnable poison rods. (orig.) [de

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
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
The calculation of Tritium burnup in Tokamaks

International Nuclear Information System (INIS)

Bittoni, E.; Haegi, M.

1987-01-01

In a deuterium plasma tokamak, the contained fusion-produced tritons are supposed to be decelerated down to thermalization according to classical Coulomb scattering. A fraction of these fast tritons undergoes the DT fusion reaction producing 14.1 MeV neutrons. It is thus possible to get information on the confinement of these fast tritons by comparing the measured and the calculated ratio of the 14.1 MeV to the 2.45 MeV neutron flux. This report describes the calculation of this flux ratio by means of a numerical Monte Carlo-like code
Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

Directory of Open Access Journals (Sweden)

Herrero J.J.

2017-01-01

Full Text Available In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.
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)
Spatially dependent burnup implementation into the nodal program based on the finite element response matrix method

International Nuclear Information System (INIS)

Yoriyaz, H.

1986-01-01

In this work a spatial burnup scheme and feedback effects has been implemented into the FERM ( 'Finite Element Response Matrix' )program. The spatially dependent neutronic parameters have been considered in three levels: zonewise calculation, assembly wise calculation and pointwise calculation. Flux and power distributions and the multiplication factor were calculated and compared with the results obtained by CITATIOn program. These comparisons showed that processing time in the Ferm code has been hundred of times shorter and no significant difference has been observed in the assembly average power distribution. (Author) [pt
ZZ ECN-BUBEBO, ECN-Petten Burnup Benchmark Book, Inventories, Afterheat

International Nuclear Information System (INIS)

Kloosterman, Jan Leen

1999-01-01

Description of program or function: Contains experimental benchmarks which can be used for the validation of burnup code systems and accompanied data libraries. Although the benchmarks presented here are thoroughly described in literature, it is in many cases not straightforward to retrieve unambiguously the correct input data and corresponding results from the benchmark Descriptions. Furthermore, results which can easily be measured, are sometimes difficult to calculate because of conversions to be made. Therefore, emphasis has been put to clarify the input of the benchmarks and to present the benchmark results in such a way that they can easily be calculated and compared. For more thorough Descriptions of the benchmarks themselves, the literature referred to here should be consulted. This benchmark book is divided in 11 chapters/files containing the following in text and tabular form: chapter 1: Introduction; chapter 2: Burnup Credit Criticality Benchmark Phase 1-B; chapter 3: Yankee-Rowe Core V Fuel Inventory Study; chapter 4: H.B. Robinson Unit 2 Fuel Inventory Study; chapter 5: Turkey Point Unit 3 Fuel Inventory Study; chapter 6: Turkey Point Unit 3 Afterheat Power Study; chapter 7: Dickens Benchmark on Fission Product Energy Release of U-235; chapter 8: Dickens Benchmark on Fission Product Energy Release of Pu-239; chapter 9: Yarnell Benchmark on Decay Heat Measurements of U-233; chapter 10: Yarnell Benchmark on Decay Heat Measurements of U-235; chapter 11: Yarnell Benchmark on Decay Heat Measurements of Pu-239
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
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)
JNC results of BN-600 benchmark calculation (phase 3)

International Nuclear Information System (INIS)

Ishikawa, M.

2002-01-01

The present work is the result of phase 3 BN-600 core benchmark problem, meaning burnup and heterogeneity. Analytical method applied consisted of: JENDL-3.2 nuclear data library, group constants (70 group, ABBN type self shielding transport factors), heterogeneous cell model for fuel and control rod, basic diffusion calculation (CITATION code), transport theory and mesh size correction (NSHEX code based on SN transport nodal method developed by JNC). Burnup and heterogeneity calculation results are presented obtained by applying both diffusion and transport approach for beginning and end of cycle
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)
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
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
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
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

pellet surface than the bulk of the pellet in leaching experiments. Thus, formation of oxidising species and radicals by radiolysis is expected to be disproportionately high as well. Therefore, when discussing high burnup fuel dissolution, the effect of the increased radiation field with burnup, as well as of the influence of the smaller grain size and increased porosity at the rim are mentioned as factors which contribute to increased dissolution rates. A third factor, increased fission product and actinide doping with burnup, has been discussed extensively in connection with increased resistance to air oxidation of the fuel. Samples from four different fuel rods, all operated in Pressurised Water Reactors (PWR), are used in the new series of corrosion experiments. They cover a burnup range from 58 to 75 MWd/kgU. The nuclide inventory of all four samples was determined by means of a combination of experimental nuclide analysis and sample specific modelling calculations. More than 40 different nuclides were analysed by isotope dilution analysis using Inductively Coupled Plasma Mass Spectrometry (ICP-MS), as well as other ICP-MS and gamma spectrometric methods. The content of roughly all fission products and actinides was also calculated separately for each sample. The experiments are performed under oxidising conditions in synthetic groundwater at ambient temperature. In order to make results as comparable as possible to those of the Series 11 experiments, the same procedure and the same leachant is used. At least nine consecutive contact periods of one and three weeks and two, three, six and twelve months are planned. The present report covers the first five contact periods up to a cumulative contact time of one year for all four samples and in addition the sixth period up to a cumulative contact time of two years for two of the samples. The samples, kept in position by a platinum wire spiral, are exposed to synthetic groundwater in a Pyrex flask. After the contact
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

pellet surface than the bulk of the pellet in leaching experiments. Thus, formation of oxidising species and radicals by radiolysis is expected to be disproportionately high as well. Therefore, when discussing high burnup fuel dissolution, the effect of the increased radiation field with burnup, as well as of the influence of the smaller grain size and increased porosity at the rim are mentioned as factors which contribute to increased dissolution rates. A third factor, increased fission product and actinide doping with burnup, has been discussed extensively in connection with increased resistance to air oxidation of the fuel. Samples from four different fuel rods, all operated in Pressurised Water Reactors (PWR), are used in the new series of corrosion experiments. They cover a burnup range from 58 to 75 MWd/kgU. The nuclide inventory of all four samples was determined by means of a combination of experimental nuclide analysis and sample specific modelling calculations. More than 40 different nuclides were analysed by isotope dilution analysis using Inductively Coupled Plasma Mass Spectrometry (ICP-MS), as well as other ICP-MS and gamma spectrometric methods. The content of roughly all fission products and actinides was also calculated separately for each sample. The experiments are performed under oxidising conditions in synthetic groundwater at ambient temperature. In order to make results as comparable as possible to those of the Series 11 experiments, the same procedure and the same leachant is used. At least nine consecutive contact periods of one and three weeks and two, three, six and twelve months are planned. The present report covers the first five contact periods up to a cumulative contact time of one year for all four samples and in addition the sixth period up to a cumulative contact time of two years for two of the samples. The samples, kept in position by a platinum wire spiral, are exposed to synthetic groundwater in a Pyrex flask. After the contact

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
Model for evolution of grain size in the rim region of high burnup UO{sub 2} fuel

Energy Technology Data Exchange (ETDEWEB)

Xiao, Hongxing, E-mail: xiaohongxing2003@163.com; Long, Chongsheng; Chen, Hongsheng

2016-04-01

The restructuring process of the high burnup structure (HBS) formation in UO{sub 2} fuel results in sub-micron size grains that accelerate the fission gas swelling, which will raise some concern over the safety of extended the nuclear fuel operation life in the reactor. A mechanistic and engineering model for evolution of grain size in the rim region of high burnup UO{sub 2} fuel based on the experimental observations of the HBS in the literature is presented. The model takes into account dislocations evolution under irradiation and the grain subdivision occur successively at increasing local burnup. It is assumed that the original driving force for subdivision of grain in the HBS of UO{sub 2} fuel is the production and accumulation of dislocation loops during irradiation. The dislocation loops can also be annealed through thermal diffusion when the temperature is high enough. The capability of this model is validated by the comparison with the experimental data of temperature threshold of subdivision, dislocation density and sub-grain size as a function of local burnup. It is shown that the calculated results of the dislocation density and subdivided grain size as a function of local burnup are in good agreement with the experimental results. - Highlights: • A model for evolution of dislocation density and grain size in HBS is proposed. • The dislocation can also be annealed when the temperature is high enough. • Original driving force for subdivision is mostly accumulation of dislocation loops. • The temperature threshold of the subdivision is predicted at 1300–1400 K.
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.
Evaluation technology for burnup and generated amount of plutonium by measurement of Xenon isotopic ratio in dissolver off-gas at reprocessing facility (Joint research)

International Nuclear Information System (INIS)

Okano, Masanori; Kuno, Takehiko; Shirouzu, Hidetomo; Yamada, Keiji; Sakai, Toshio; Takahashi, Ichiro; Charlton, William S.; Wells, Cyndi A.; Hemberger, Philip H.

2006-12-01

The amount of Pu in the spent fuel was evaluated from Xe isotopic ratio in off-gas in reprocessing facility, is related to burnup. Six batches of dissolver off-gas (DOG) at spent fuel dissolution process were sampled from the main stack in Tokai Reprocessing Plant (TRP) during BWR fuel (approx. 30GWD/MTU) reprocessing campaign. Xenon isotopic ratio was determined with Gas Chromatography/Mass Spectrometry. Burnup and generated amount of Pu were evaluated with Noble Gas Environmental Monitoring Application code (NOVA), developed by Los Alamos National Laboratory. Inferred burnup evaluated by Xe isotopic measurements and NOVA were in good agreement with those of the declared burnup in the range from -3.8% to 7.1%. Also, the inferred amount of Pu in spent fuel was in good agreed with those of the declared amount of Pu calculated by ORIGEN code in the range from -0.9% to 4.7%. The evaluation technique is applicable for both burnup credit to achieve efficient criticality safety control and a new measurement method for safeguards inspection. (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)
Determination of burnup, cooling time and initial enrichment of PWR spent fuel by use of gamma-ray activity ratios

International Nuclear Information System (INIS)

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

1999-01-01

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

International Nuclear Information System (INIS)

Maeda, Shigetaka; Sekine, Takashi; Kitano, Akihiro; Nagasaki, Hideaki

2005-03-01

The MK-III performance test began in June 2003 to fully characterize the upgraded core and heat transfer system of the experimental fast reactor JOYO. This paper describes the results of the approach to criticality, the excess reactivity evaluation and the burn-up coefficient measurement. In the approach to criticality test, the MK-III core achieved initial criticality at the control rod bank position of 412.8 mm on 14:03 July 2nd, 2003. Because the replacement of the outer two rows of reflector subassemblies with shielding subassemblies reduced the source range monitor signals by a factor of 3 at the same reactor power compared with those in the MK-II core, we measured the change of the monitor's response and determined the count rate 2x10 4 cps.' as an appropriate value judging the zero power criticality. In the excess reactivity evaluation, the zero power excess reactivity at 250degC was 2.99±0.10%Δk/kk' based on the measured critical rod bank position and the measured control rod worths. The predicted value by the JOYO core management code system HESTIA was 3.13±0.16%Δk/kk', showing good agreement with the measured value. The measured excess reactivity was within the safety requirement limit. In the burn-up coefficient measurement, the excess reactivity change versus the reactor burn-up was evaluated. The measurement method adopted was to measure the control rod positions during the rated power operation. A value of -2.12x10 -4 Δk/kk'/MWd was obtained as a measured burn-up coefficient. The value calculated by HESTIA was -2.12x10 -4 Δk/kk'/MWd, and it agreed well with the measured value. All technical safety requirements for MK-III core were satisfied and the calculation accuracy of the core management code system HESTIA was confirmed. (author)
ZZ PWR-AXBUPRO-GKN, Measured Axial Burnup Profiles, NPP Neckarewstheim

International Nuclear Information System (INIS)

Neuber, Jens-Christian; Lamprecht, Thomas

1999-01-01

Description or function: PWR-AXBUPRO-GKN12 contains Axial Burnup Shapes released by Siemens AG Power Generation Group. It contains data sets relative to following NPP and initial enrichment: - NPP Neckarwestheim 1, Fuel assemblies with an initial enrichment of 3.5 wt.-% 235-U; - NPP Neckarwestheim 2, Fuel assemblies with an initial enrichment of 3.5 wt.-% 235-U; - NPP Neckarwestheim 2, Fuel assemblies with an initial enrichment of 3.8 wt.-% 235-U; - NPP Neckarwestheim 2, Fuel assemblies with an initial enrichment of 4.0 wt.-% 235-U. In each of these files the axial shapes are listed one after the other. Each shape is characterised by: - the number of the cycle; - the number of the fuel assembly. - axial burnup shape characteristics: - height (in centimeter) of the nodes with respect to the active length of the fuel assemblies normalized to the cold, unirradiated state. - the nodal burnup (in MWd/kg U). - fuel assembly design data as well as core geometry and operating data pertinent to depletion calculations: NPP Neckarwestheim 1 (GKN1) - (square pitch lattice 15X15 - thermal Power 2497 MW)thermal Power 2497 MW); More than 700 EOC axial shapes from cycle 18 up. From cycle 18 to cycle 20 a change from an Out-In-Loading to an In-Out-Loading has taken place. Fuel assemblies up to number 1093 have spacer grids made of Inconel, whereas all the fuel assemblies from number 1094 up have spacer grids made of Zircaloy. Discharge burnups range from: 9.7 to 52.8 MWd/kg. NPP Neckarwestheim 2 (GKN2) - (square pitch lattice 18X18 - thermal Power 3850 MW) more than 500 EOC axial shapes from cycle 5 up: - More than 170 shapes for an initial fuel enrichment of 3.5 wt.-% 235-U, discharge burnup ranges from 16.3 to 44.4 MWd/kg; - more than 170 shapes for an initial fuel enrichment of 3.8 wt.-% 235-U, discharge burnup ranges from 14.0 to 52.8 MWd/kg; - more than 180 shapes for an initial fuel enrichment of 4.0 wt.-% 235-U. discharge burnup ranges from 15.5 to 48.9 MWd/kg. PWR AXBUPRO
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
DRAGON, Reactor Cell Calculation System with Burnup

International Nuclear Information System (INIS)

2007-01-01

1 - Description of program or function: DRAGON is a collection of models to simulate the neutronic behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations which can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. The user must supply cross sections. DRAGON can access directly standard microscopic cross-section libraries in the following formats: DRAGON, MATXS (TRANSX-CTR), WIMSD4, WIMS-AECL, and APOLLO. It has the capability of exchanging macroscopic and microscopic cross-section libraries with a code such as PSR-0206/TRANSX-CTR or PSR-0317/TRANSX-2 by the use of the GOXS and ISOTXS format files. Macroscopic cross sections can also be read in DRAGON via the input data stream. 2 - Method of solution: DRAGON contains a multigroup iterator conceived to control a number of different algorithms for the solution of the neutron transport equation. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are included in a source term. The current version, DRAGON 9 71124 (Release 3.02), which was released in January 1998, contains three such algorithms. The JPM option solves the integral transport equation using the interface current method applied to homogeneous blocks; the SYBIL option solves the integral transport equation using the collision probability method for simple one-dimensional (1-D) or two-dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; and the
Numerical solution of the point reactor kinetics equations with fuel burn-up and temperature feedback

International Nuclear Information System (INIS)

Tashakor, S.; Jahanfarnia, G.; Hashemi-Tilehnoee, M.

2010-01-01

Point reactor kinetics equations are solved numerically using one group of delayed neutrons and with fuel burn-up and temperature feedback included. To calculate the fraction of one-group delayed neutrons, a group of differential equations are solved by an implicit time method. Using point reactor kinetics equations, changes in mean neutrons density, temperature, and reactivity are calculated in different times during the reactor operation. The variation of reactivity, temperature, and maximum power with time are compared with the predictions by other methods.
Improved and consistent determination of the nuclear inventory of spent PWR-fuel on the basis of time-dependent cell-calculations with KORIGEN

International Nuclear Information System (INIS)

Fischer, U.; Wiese, H.W.

1983-01-01

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

International Nuclear Information System (INIS)

Spasojevic, D.; Matausek, M.; Marinkovic, N.

1989-01-01

Prolonged nuclear fuel burnup in advanced NPP with four or more instead of three one-year cycles, and/or with 15- to 18-month instead of standard 12-month cycles, requires the fresh fuel to have increased enrichment combined with burnable poisons. This causes changes in axial and radial distribution of power generation during the particular fuel cycles, so that detailed analyses of thermal reliability of reactor core becomes necessary. This paper presents the results of the analysis of the departure from nuclear boiling ratio DNBR for an equilibrium cycle of an advanced PWR. (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
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.
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.)
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
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)
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)
FRAPCON-3: A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup

International Nuclear Information System (INIS)

Berna, G.A.; Beyer, G.A.; Davis, K.L.; Lanning, D.D.

1997-12-01

FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2) cladding elastic and plastic deformation, (3) fuel-cladding mechanical interaction, (4) fission gas release, (5) fuel rod internal gas pressure, (6) heat transfer between fuel and cladding, (7) cladding oxidation, and (8) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat-transfer correlations. The codes' integral predictions of mechanical behavior have not been assessed against a data base, e.g., cladding strain or failure data. Therefore, it is recommended that the code not be used for analyses of cladding stress or strain. FRAPCON-3 is programmed for use on both mainframe computers and UNIX-based workstations such as DEC 5000 or SUN Sparcstation 10. It is also programmed for personal computers with FORTRAN compiler software and at least 8 to 10 megabytes of random access memory (RAM). The FRAPCON-3 code is designed to generate initial conditions for transient fuel rod analysis by the FRAPTRAN computer code (formerly named FRAP-T6)
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
Analysis of UO2 fuel structure for low and high burn-up and its impact on fission gas release

International Nuclear Information System (INIS)

Szuta, M.; El-Koliel, M.S.

1999-01-01

During irradiation, uranium dioxide (UO 2 ) fuel undergo important restructuring mainly represented by densification and swelling, void migration, equiaxed grain growth, grain subdivision, and the formation of columnar grains. The purpose of this study is to obtain a comprehensive picture of the phenomenon of equiaxed grain growth in UO 2 ceramic material. The change of the grain size in high-density uranium dioxide as a function of temperature, initial grain size, time, and burnup is calculated. Algorithm of fission gas release from UO 2 fuel during high temperature irradiation at high burnup taking into account grain growth effect is presented. Theoretical results are compared with experimental data. (author)
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.
Determination of burnup for IEAR-1 fuel elements by non destructive method of gamma spectrometry

International Nuclear Information System (INIS)

Madi Filho, T.; Holland, L.

1982-01-01

Burnup determination, by non-destructive gamma spectrometry of spent fuel with high and low activity of IEAR-1 reactor, using Cs-137 as burnup monitor, were done. To measure the Cs-137 distribution in these elements a Ge(Li) detector, with volume equal to 73,7 cm 3 , in two measurement systems with defined geometry and good colimation, was used. The IEA-14 taken from the core about 20 years ago, presents a gamma spectra due to Cs-137. The IEA-80, with cooling time approximately to 5 years, shows a more complex gamma spectrum due to other fission products still found in significant quantities. The IEA-14 measures were done in a measurement system used outside the reactor pool (S.I.), being the global efficiency of this system obtained by using a plane, calibrated and extense Ag-110 m source. Detailed measures of gamma transmission, using Cs-137 as a calibrated and punctiforme source, showed the high homogenity of the fuel plates. (E.G.) [pt
The role of Monte Carlo burnup calculations in quantifying plutonium mass in spent fuel assemblies with non-destructive assay

Energy Technology Data Exchange (ETDEWEB)

Galloway, Jack D.; Tobin, Stephen J.; Trellue, Holly R.; Fensin, Michael L. [Los Alamos National Laboratory, Los Alamos, (United States)

2011-12-15

The Next Generation Safeguards Initiate (NGSI) of the United States Department of Energy has funded a multi-laboratory/university collaboration to quantify plutonium content in spent fuel (SF) with non-destructive assay (NDA) techniques and quantify the capability of these NDA techniques to detect pin diversions from SF assemblies. The first Monte Carlo based spent fuel library (SFL) developed for the NGSI program contained information for 64 different types of SF assemblies (four initial enrichments, burnups, and cooling times). The maximum amount of fission products allowed to still model a 17x17 Westinghouse pressurized water reactor (PWR) fuel assembly with four regions per fuel pin was modelled. The number of fission products tracked was limited by the available memory. Studies have since indicated that additional fission product inclusion and asymmetric burning of the assembly is desired. Thus, an updated SFL has been developed using an enhanced version of MCNPX, more powerful computing resources, and the Monte Carlo-based burnup code Monteburns, which links MCNPX to a depletion code and models a representative 1 Division-Slash 8 core geometry containing one region per fuel pin in the assemblies of interest, including a majority of the fission products with available cross sections. Often in safeguards, the limiting factor in the accuracy of NDA instruments is the quality of the working standard used in calibration. In the case of SF this is anticipated to also be true, particularly for several of the neutron techniques. The fissile isotopes of interest are co-mingled with neutron absorbers that alter the measured count rate. This paper will quantify how well working standards can be generated for PWR spent fuel assemblies and also describe the spatial plutonium distribution across an assembly. More specifically we will demonstrate how Monte Carlo gamma measurement simulations and a Monte Carlo burnup code can be used to characterize the emitted gamma
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
Carmen system: a code block for neutronic PWR calculation by diffusion theory with spacedependent feedback effects

International Nuclear Information System (INIS)

Ahnert, C.; Aragones, J.M.

1982-01-01

The Carmen code (theory and user's manual) is described. This code for assembly and core calculations uses diffusion theory (Citation), with feedback in the cross sections by zone due to the effects of burnup, water density, fuel temperature, Xenon and Samarium. The burnup calculation of a full cycle is solved in only an execution of Carmen, and in a reduced computer time. (auth.)
Parameterized representation of macroscopic cross section in the PWR fuel element considering burn-up cycles

International Nuclear Information System (INIS)

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

2015-01-01

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

International Nuclear Information System (INIS)

Tsai, Chiung-Wen

2013-01-01

The goal of this technical note is to provide an alternative, the freeware Scilab, by which means we may construct custom GUIs and distribute them without extra constrains and cost. A post-processor has been constructed by Scilab to visualize the fuel rod burnup analysis data calculated by FRAPCON-3.4. This post-processor incorporates a graphical user interface (GUI), providing users a rapid overview of the characteristics of the numerical results with 2-D and 3-D graphs, as well as the animations of fuel temperature distribution. An assessment case input file provided by FRAPCON user group was applied to demonstrate the construction of a post-processor with GUI by object-oriented GUI tool, as well as the capability of visualization functions of Scilab
Visualization of fuel rod burnup analysis by Scilab

Energy Technology Data Exchange (ETDEWEB)

Tsai, Chiung-Wen, E-mail: d937121@oz.nthu.edu.tw

2013-12-15

The goal of this technical note is to provide an alternative, the freeware Scilab, by which means we may construct custom GUIs and distribute them without extra constrains and cost. A post-processor has been constructed by Scilab to visualize the fuel rod burnup analysis data calculated by FRAPCON-3.4. This post-processor incorporates a graphical user interface (GUI), providing users a rapid overview of the characteristics of the numerical results with 2-D and 3-D graphs, as well as the animations of fuel temperature distribution. An assessment case input file provided by FRAPCON user group was applied to demonstrate the construction of a post-processor with GUI by object-oriented GUI tool, as well as the capability of visualization functions of Scilab.

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