1
Burnup and Plutonium Distribution of WWER-440 Fuel Pin
At Extended Burnup
Üner Çolak
Department of Nuclear Engineering,
Hacettepe University, Beytepe 06532 Ankara/Turkey
Tahir Akbaş, Ömer Gündüz
Nuclear Safety Department,
Turkish Atomic Energy Authority, Ankara/Turkey
The formation of rim region in LWR UO2 based nuclear fuel at high burnup is a common observation.
This region has very high porosity due to excessive gas release. Such a region is also characterized by significantly high plutonium concentration and high local burnup compared to the internal fuel region. Spatial distribution of these parameters has been incorporated with fuel behavior and performance analysis codes by using mostly empirical relations. Variation of these parameters depends on the neutron flux as well as neutron energy spectrum. Detailed neutronics analysis is necessary for the accurate prediction of these parameters. This study is performed by MCNP4B Monte Carlo code for the calculation of local neutron flux, ORIGEN2 for burnup and depletion calculations, and MONTEBURNS for coupling these codes. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell. Fuel pin is divided into a number of radial segments. A relatively small mesh size is used at the region near the surface to reveal the rim effect. The variation of plutonium and local burnup are obtained for high burnup. Results are compared with existing experimental observations for WWER-440 fuel and other theoretical predictions.
Keywords: Rim effect, WWER-440 Fuel, ORIGEN2, MCNP
1This paper was presented in the “4th International Seminar on WWER Fuel Performance, Modelling and Experimental Support”, 1-5 October 2001, Albena, Bulgaria.
1. INTRODUCTION
The tendency to increase discharge burnup beyond 60 MWd/kgHM for LWR fuel has required the development of new methodologies and models for the prediction of fuel behavior at high burnup [1,2,3]. New methodologies are especially important in the safety assessment of LWRs in fuel related aspects at extended burnup. Observations show that fuel behavior at high burnup cannot be easily modeled with traditional approaches. More detailed and accurate characterization of the distribution of heavy nuclides and burnup is an important task in fuel behavior modeling.
One of the most common observations at high burn-ups is the formation of rim region in UO2
based LWR fuels at high burnups (above 45 MWd/kgHM). The rim region consists of development of large secondary grains and extensive intra-granular porosity close to pellet outer rim. High local burnup is also observed due to the plutonium build-up in this region. This region has a very high porosity and excessive gas release.
Prediction of plutonium concentration and burnup profiles in an irradiated fuel pin is important for performance analysis of fuel during irradiation and storage after discharge. Such profiles depend on the local neutron flux as well as neutron energy spectrum. Detailed neutronics analysis is necessary for the accurate prediction of these parameters. There are several studies to calculate such parameters and to assess high burnup effects for WWER reactor fuels [4,5,6]. The main objective of this study is to generate radial profiles of plutonium distribution and local burnup for a WWER fuel pin at high burnup, and to assess the results in the view of relevant data available in literature.
2. WWER-440 FUEL AND PIN CELL MODEL
A representative WWER-440 fuel rod is selected for the analysis. A hexagonal pin cell is constructed as shown in Figure 1. Fuel rod is irradiated to a burnup of ≈70 MWD/kgHM. WWER-440 fuel pin parameters used in the study are given in Table 1. Fuel pin is divided into 40 concentric cylindrical shells with equal volumes to be able to predict radial plutonium profile accurately, especially in the rim region. The linear heat generation rate of 325 W/cm is selected to represent high burnup region for a WWER-440 fuel rod. In the model, 2 cm axial length of the fuel rod is considered to avoid axial variation of linear heat generation rate.
3. METHODOLOGY
In this work, MCNP4B and ORIGEN2 are utilized for the calculation of local neutronics parameters and depletion characteristics, respectively. The data transfer and file handling operations are facilitated by MONTEBURNS code.
MCNP [7] is a neutral particle and electron transport code that uses the Monte Carlo technique. Monte Carlo technique considers the random nature of the processes. Each process is characterized by a probability distribution function. The transport of particles in a medium is followed and a history is generated for each particle. Interactions between the particle and host atoms are determined by cross section values and probability distribution functions. All kinds of interactions, elastic and inelastic scattering, capture, and fission, are considered as possible modes of interaction. In order to obtain reasonable and accurate results with a Monte Carlo simulation, one needs to repeat simulations for a large number of particles. Thus, a standard deviation associated with the mean will show the simulation performance. Particle flux at any given point can be obtained by a simple bookkeeping operation. Particle flux values are then used to obtain a wide
variety of information such as reaction rates, radiation dose, energy deposition and so on. Another application of Monte Carlo method in MCNP is the criticality analysis. This analysis is done by a simple balance operation. The number of neutron in the system and their interactions as well as leakages is considered and effective multiplication factor is evaluated by MCNP. Even though MCNP is a versatile computer code to analyze many nuclear systems, it is not capable of making burnup and depletion calculations.
ORIGEN [8] is a computer code system for calculating the buildup, decay, and processing of radioactive materials and developed by Oak Ridge National Laboratory in the late 1960s and early 1970s. ORIGEN2 is a revised version of ORIGEN and incorporates updates of the reactor models, cross sections, fission product yields, decay data, and decay photon data, as well as the source code. ORIGEN2.1 that is used in this study replaces ORIGEN2 and includes additional libraries for standard and extended burnup PWR and BWR calculations.
Computer codes MCNP and ORIGEN are sufficient for the analysis aimed in this study. These codes produce necessary cross information to be used as input parameters. However, it is very cumbersome to extract useful data from output files and prepare necessary input files during many stages of computations. Another difficulty is the file management. These difficulties are overcome by the use of the utility code MONTEBURNS [9].
MONTEBURNS, which links Monte Carlo transport code MCNP4B with fuel depletion code ORIGEN2.1, is used in this study. MONTEBURNS consists of a Perl script file that frequently interacts with MONTEB, a FORTRAN77 code. A number of calculations are performed by the code for each time step, such as energy per fission, flux normalization, reactor physics constants, effective multiplication factor, power, fractional importance, and activity, heat load and radio toxicities. MONTEBURNS is designed to link the Monte Carlo transport code MCNP with the radioactive decay and depletion code ORIGEN2. MONTEBURNS produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. The program processes input from the user that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Various results from MCNP, ORIGEN2, and other calculations are then output successively as the code runs. MONTEBURNS performs one or more MCNP and ORIGEN2 runs for each specified time step.
An MCNP model is prepared for this pin cell and depletion information is given in MONTEBURNS input file. The number of particles to be simulated is selected to have relative error below the acceptable levels (0.05) in Monte Carlo calculations. One of input parameters of MCNP is number of isotopes to be tallied. If an isotope contributes a large enough fraction (i.e., greater than fractional importance) to absorption, fission interactions, or atom density, then, this isotope is included to the isotopes to be tallied in MCNP run. Fractional importance is 0.0001 in this study. Small fractional importance will result in more accurate results and unfortunately will require more computer time.
Pin cell model is used in MCNP4B with appropriate boundary conditions. Surfaces of hexagon are characterized by periodic surfaces. Hence, infinite array of fuel pins are represented in the model. Similarly, top and bottom surfaces of the geometric model are represented by reflected surfaces so that neutron leakage from these surfaces is avoided. ORIGEN2 performs burnup calculations for MONTEBURNS using the matrix exponential method. ORIGEN2 considers time-dependent formation, destruction, and decay concurrently. These calculations require (1) the initial compositions and amounts of material, (2) one-group microscopic cross-sections for each isotope,
(3) material feed and removal rates (if desired), (4) the length of the irradiation period(s), and (5) the flux or power of the irradiation. Input files for ORIGEN2 are complicated to write, and output files generated by ORIGEN2 are bulky and complicated to read. Thus, MONTEBURNS eliminates the user’s need to create his/her own ORIGEN2 input files and to extract information from ORIGEN2 output files.
4. RESULTS and DISCUSSION
Radial local burnup distribution obtained at various pin average burnup levels is given in Figure 2. The burnup in the rim region (rrel>0.95) increases up to almost two times of pin average burnup.
This profile is mainly because of the higher fission rate at the peripheral region of fuel pin as expected. Local burnup distribution profile is in good agreement with results of other studies performed for WWER-440 fuel rods [5,6].
Radial plutonium distribution at various pin average burnup levels is given in Figure 3. The Pu generated in an irradiated UO2 fuel pin is distributed non-uniformly in radial direction. Formation
of Pu is proportional to fission rate and local conversion ratio (Figure 4). Conversion ratio is defined as the ratio of production rate of fissile atoms to the consumption rate of fissile atoms. Since the rim region of the fuel pin has higher fission rate and conversion ratio, Pu weight percent in this region is two or three times greater than that of in the central region of the pin at high average burnup levels.
At the maximum average burnup (70 MWd/kgHM) the Pu concentration is 1.23 w% near the central hole and 3.4 w% near the surface of the pin. It is reported that Pu concentration determined by mass-spectrometric methods near to central hole and the in the rim region is 1.5 w% and 4.2 w% respectively for high burnup WWER-440 fuel rods [10]. Rim region Pu mass fraction is reported as 3-4 w% in another experimental study [11]. Pu distribution calculated in this study is in good agreement with these results.
Radial distribution profiles of Xenon and Krypton as the principles components of the gases in the fuel rod are also important in fuel analysis. Xenon and Krypton radial distribution profiles are given in Figure 5. These results are relatively different from the results obtained by codes having mechanical and physical models such as ASFAD[5] and TRANSURANUS [12].
5. CONCLUSION
Burnup and Pu distribution profiles at high burnup WWER-440 fuel pin are determined by MONTEBURNS that couples MCNP and ORIGEN2. Burnup and Pu distributions are in good agreements with experimental results and other theoretical predictions.
It is possible to increase accuracy of calculations. To get more accurate results, the following may be suggested:
! To break up irradiation period into the shorter time periods because the physics and composition of materials in the system may change significantly during the given time interval,
! To increase the number of MCNP/ORIGEN2 runs per time step, or ! To decrease the fractional importance.
Since complex processes taking place in fuel elements such as actinide redistribution, atomic diffusion, gas release and so on are not included in the calculations, it is not expected to obtain similar results for volatile fission products with the experimental studies. However, accurate calculations performed with codes utilized in this study may be used to provide information in terms of local burnup and plutonium concentration in order to improve materials property assessment, which is important in fuel performance analysis.
6. REFERENCES
[1] J. Secker, et al. “Long Cycle and High Burnup Fuel Experience at Westinghouse”, 3rd International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, 4-8 October 1999, Pamporovo, Bulgaria.
[2] A. Grishakov, et al. “The Ways of WWER-440 Reactor Assemblies and Fuel Cycles Design
Improvement”, 3rd International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, 4-8 October 1999, Pamporovo, Bulgaria.
[3] V. Molchanov, et al. “Status and Trends of WWER Fuel and Fuel Cycles”, 3rd International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, 4-8 October 1999, Pamporovo, Bulgaria.
[4] M. Solonin, Yu., et al. “WWER Fuel Performance and Material Devolopment for Extended
Burnup in Russia”, 2nd International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, 21-25 April 1997, Sandanski, Bulgaria.
[5] S. Lemehov, “ASFAD Code Calculations for WWER-440 High Burnup fuel Rods”, 2nd International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, 21-25 April 1997, Sandanski, Bulgaria.
[6] P. Strizhov, B. Volkov. “Integration of PIN and FRASM Codes Improved WWER Fuel
Simulation under Operational and Accident Conditions”, 3rd International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, 4-8 October 1999, Pamporovo, Bulgaria.
[7] J. F. Briesmeister, “MCNP – A General Monte Carlo N-Particle Transport Code, Version 4B”, Los Alamos National Laboratory Document LA-12625-M, March 1997.
[8] RSICC Computer Code Collection, “ORIGEN 2.1: Isotope Generation and Depletion Code,
Matrix Exponential Method”, Oak Ridge National Laboratory Document CCC-371, December
1991.
[9] D. I. Poston, H. R. Trellue, “Monteburns Version 1.0”, Los Alamos National Laboratory Document LA-UR-99-4999, September 1999.
[10] A. Smirnov, et al. “Behavior of WWER-440 and WWER-1000 fuel in a Burnup Range of
20-48 MWd/kgU”, 2nd International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, 21-25 April 1997, Sandanski, Bulgaria.
[11] A. Smirnov, et al. “The Peculiarities of the WWER-440 Fuel Behavior at Higher Burnups”, 2nd International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, 21-25 April 1997, Sandanski, Bulgaria.
[12] M. Manalova, et al. “Validation of TRANSURANUS Code-WWER Version by the Updated
KOLA-3 Data Set”, 3rd International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, 4-8 October 1999, Pamporovo, Bulgaria.
Table 1. Parameters of WWER-440 Unit Cell
Parameter Value
Fuel Properties
Clad material Zr-1%Nb Clad outer diameter 0.91 cm Clad inner diameter 0.78 cm
Filling gas He
Fuel pin material UO2
Fuel density 10.6 gr/cm3 Fuel pin outer diameter 0.757 cm Fuel pin hole diameter 0.15 cm Pitch size 1.22 cm
Irradiation History
Fuel enrichment 4.4% Linear Heat Generation Rate 325 W/cm Irradiation time 900 days
Figure 1. WWER-440 Fuel Cell
Burnup Distribution Profile at Varius Average Burnup Levels
0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 110.0 120.0 130.0 140.0 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.90 1.00
Relative Pellet Radius
Burnup (MWd/kgHM)
24 47 70
Pu Distribution Profile at Varius Average Burnup Levels 0.50 1.00 1.50 2.00 2.50 3.00 3.50 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.90 1.00
Relative Pellet Radius
Pu (w%)
24 47 70
Figure 3. Radial Pu Distribution
Radial Conversion Ratio Profile (12 MWd/kgHM)
0.00 0.20 0.40 0.60 0.80 1.00 1.20 1.40 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.90 1.00
Relative Pellet Radius
Conversion Ratio
Radial Xe Distribution Profile at Varius Average Burnup Levels 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.00 0.20 0.40 0.60 0.80 1.00
Relative Pellet Radius
Xe (w%)
24 47 70
Figure 5. Radial Xenon Distribution
Radial Kr Distribution Profile at Varius Average Burnup Levels
0.02 0.03 0.04 0.05 0.06 0.07 0.08 0.09 0.1 0.11 0.12 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.90 1.00
Relative Pellet Radius
Kr (w%)
24 47 70
