Comparison of the results of calculations and post-reactor investigations of VVÉR-1000 fuel elements with burnup 49 MW·days/kg

The results of calculations performed with the PINw99, TRANSURANUS (V1M1V03), and TOPRA-2 computer programs are compared with data obtained from post-reactor investigations of fuel elements which operated for four years in the No. 1 unit of the Zaparozh’e nuclear power plant with a VVéR-1000 reactor to burnup ≈ 49 MW·days/kg. The initial data are analyzed, and a comparison is made of the computed and experimental elongation of the fuel elements (49 fuel elements), the yield of gaseous fission products and the subcladding pressure (35 fuel elements), and the decrease of cladding diameter and fuel-cladding gap width. It is shown that these computer programs can be used to calculate VVéR fuel elements. __________ Translated from Atomnaya énergiya, Vol. 101, No. 6, pp. 413–420, December, 2006.     

Interaction of uranium-molybdenum fuel with an aluminum matrix with deep burnup

A thermodynamic analysis is used to calculate the phase and component composition of uraniummolybdenum fuel with burnup 200 GW·days/ton. The equilibrium composition of the gas phase, consisting mainly of gaseous cesium whose pressure reaches 30 kPa, is determined more accurately. The quantitative composition of the phase of solid solutions of tellurides, whose formation degrades the structure of a fuel granule, is presented. Thermal tests of the fuel composition (U–Mo)–Al were performed. The investigation was performed in the presence of simulators of the chemically active fission products of cesium and iodine at different temperature. The interaction zone of (U–Mo)–Al is investigated by means of metallography and scanning electron microscopy. The data obtained on the composition of the indicated zone made it possible to conjecture the character of interaction between the fuel material and the aluminum matrix.     

Quantitative accuracy analysis of burnup calculations for BNPP fuel assemblies using FFTBM method

A Fast Fourier Transform Based Method (FFTBM) analysis is carried out to determine the accuracy and validation of burnup calculations for Bushehr Nuclear Power Plant (BNPP), with two different codes. WIMSD5 and ORIGEN-2 are used for burnup calculations. Two types of fuel assembly configurations are modeled in evaluating 14 fission products' (FPs) concentrations during 2 and 3 year fuel cycles. The fission products are selected based on their radiological importance in shielding, the energy of gamma radiation, half life, availability of isotope from WIMS and ORIGEN results and their importance in criticality calculations.Despite different approach of burnup calculations by the two codes, FFTBM quantitative analysis showed very good agreement between the results.     

Investigation of fuel structure and composition after irradiation in VVÉR-440 to burnup 70 MW·days/kg

The results of structural investigations performed on fuel and fission products — neodymium, xenon, and cesium — along the radius of a fuel kernel after irradiation in VVéR-440 to burnup 70.2 MW·days/kg are presented. The radial distribution of neodymium is used to calculate the radial distribution of burnup and the accumulation of xenon and cesium. It is shown that a decrease of the xenon content in the fuel matrix as compared with the amount formed over the irradiation time is observed over the entire cross section of the pellet and is due to complete or partial fuel recrystallization occurring predominately along the boundaries of the initial grains and characterized by the formation of a fine-grain structure together with submicron and micron pores. __________ Translated from Atomnaya énergiya, Vol. 101, No. 4, pp. 286–289, October, 2006.     

Experimental-computational investigation of the energy-release field and field burnup in IRT

A comprehensive γ-spectrometric investigation of fuel assemblies in an IRT core was performed. The objective was to obtain experimental information on the contribution of individual fuel assemblies to the total energy production and the relative fuel burnup and to compare the results with the computed prediction obtained with the TIGR program. Measurements of the relative burnup and energy production (in the last 4 months) of individual fuel assemblies in the IRT core were performed. The error in the obtained data is 2–3%, which permits using them as tests for calculations. It was established that the accuracy of the calculations of the relative contribution of a fuel assembly to the energy production and burnup, performed using the TIGR program, is satisfactory, 5 figures, 3 tables, 4 references. Moscow Engineering Physics Institute. Translated from Atomnaya énergiya, Vol. 88, No. 6, pp. 426–431, June, 2000.     

Chemical state of fission products in irradiated uranium carbide fuel

The chemical state of fission products in irradiated uranium carbide fuel has been estimated by equilibrium calculation using the SOLGASMIX-PV program. Solid state fission products are distributed to the fuel matrix, ternary compounds, carbides of fission products and intermetallic compounds among the condensed phases appearing in the irradiated uranium carbide fuel. The chemical forms are influenced by burnup as well as stoichiometry of the fuel. The results of the present study almost agree with the experimental ones reported for burnup simulated carbides.     

Numerical solution of the chain model of fission product nuclides. Effect on the reactivity of the reactor

Runge–Kutta algorithm of fourth order devised for the numerical solution of ordinary differential equations of fuel isotopic composition and fission products accumulation systems is described. Several hundreds fission product isotopes are included in partially depleted fuel as a result of direct yield from nuclear fission reaction, radioactive decay of fission products and/or neutron capture. The algorithm of the suggested method was coded and compared with both the analytical method and the international ORIGEN2 code system. The accuracy and speed of the calculations as obtained by implementing the method in a PCs with Linux system are found to be acceptable and faster than their resemblance. The effect of the calculated fission product concentrations on the reactivity of the reactor has been included.     

Criticality of the thorium burnup in equilibrium state

The effective neutron multiplication factor (k_eff) as a function of burnup for different volume coolant (CoR) and fuel (FR) to cell ratio is presented. Additionally the Conversion Ratio (CR) of Th-232 to U-233, concentration of U-233, fissile and fission products calculation as a function of burnup are presented. The assembly is a critical reactor which makes volumes of coolant and fuel changes possible. In addition, an analytical model of calculation of k_eff as a function of U-233 and a poison concentration in equilibrium state are presented. One can achieve the criticality of Thorium Breeder Reactor (TBR) for enough high average neutron energy which one can obtain in Fast Breeder Reactor (FBR) only. The maximal value of CR and burnup for case of k_eff ≥ 1 achieves 1.4 and 360 GWd/MTU, correspondently. The calculations were done with a MCNPX 2.7 code using F₂Be, Na and Pb coolants.     

基于CASMO5的燃耗历史对乏燃料反应性的影响计算

基于乏燃料贮存领域常用的锕系加裂变产物(APU-2)级燃耗信任制,应用二维组件燃耗计算程序CASMO5,计算了燃耗过程中功率密度和运行历史对乏燃料k∞的影响。结果表明:燃耗计算中,选择堆芯额定功率对应的平均功率密度,同时k∞附加0.002 3的包络裕度,运行历史选择循环内及循环间无停堆额定功率运行,同时k∞附加0.004 5的包络裕度,可满足燃耗信任制中包络性原则。     

MCNPX Monte Carlo burnup simulations of the isotope correlation experiments in the NPP Obrigheim

This paper describes the simulation work of the Isotope Correlation Experiment (ICE) using the MCNPX Monte Carlo computer code package. The Monte Carlo simulation results are compared with the ICE-Experimental measurements for burnup up to 30 GWD/t. The comparison shows the good capabilities of the MCNPX computer code package for predicting the depletion of the uranium fuel and the buildup of the plutonium isotopes in a PWR thermal reactor. The Monte Carlo simulation results show also good agreements with the experimental data for calculating several long-lived and stable fission products. However, for the americium and curium actinides, it is difficult to judge the predication capabilities for these actinides due to the large uncertainties in the ICE-Experimental data. In the MCNPX numerical simulations, a pin cell model is utilized to simulate the fuel lattice of the nuclear power reactor. Temperature dependent libraries based on JEFF3.1 nuclear data files are utilized for the calculations. In addition, temperature dependent libraries based ENDF/B-VII nuclear data files are utilized and the obtained results are very close to the JEFF3.1 results, except for ∼10% differences in the prediction of the minor actinide isotopes buildup.     

Analysis of relative release rates of radionuclides from the RBMK-1500 reactor fuel elements

Specific activities (concentrations) of fission products (FP) and activation products in spent fuel elements of the RBMK-1500 reactor were calculated using SCALE 5 computer code. Different burnup (5.1–21.0 MWd/kg) fuel assemblies were experimentally investigated. Activities of radionuclides present in the coolant water of storage cases of defective fuel elements were experimentally measured and analyzed. Experimental results provide a basis for a quantitative analysis of radionuclide release from spent fuel of the RBMK-1500 reactor. Relative release rates of radionuclides from the fuel matrix were assessed based on a comparison of experimental results with theoretical calculations. On the basis of analysis results released fission and activation products can be divided into several groups according to their release rates from fuel; this can be generalized for radionuclides with similar chemical properties.     

Simulation of high-temperature swelling of uranium dioxide and deformation behavior of a fuel element

Models and computer codes, developed based on them, for simulating the swelling of uranium dioxide (BARS) and the stress-deformation state of a fuel element (SDS) under high-temperature irradiation are presented. It is shown that when developing a design for high-temperature fuel elements and validating their serviceability the quantitative indicator required for the swelling of uranium dioxide in the range ≥1400°C is the change in the external dimensions of the fuel caused by constant formation and growth of bubbles containing gaseous fission products during irradiation. The results of computational investigations using the models indicated are examined. These results eliminate the inconsistency of the data on the effect of the main operating parameters — the temperature and burnup — on the radiation characteristics and service life behavior of a fuel element. It is shown that the central channel in the fuel kernel and strengthening of the cladding improve the dimensional stability fuel elements. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 172–179, September, 2007.     

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO₂ or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.     

Verification of the BONUS module integrated into the SOCRAT code

The results of a verification of the BONUS method as regards predictions of the time dependence of the mass and activity of fission products produced in thermal reactors are presented. Different standard regimes of fuel irradiation in VVER-1000 are examined, and the results calculated using the autonomous version of the BONUS code and as a module integrated into SOCRAT code are compared with the results obtained using other codes, including precision program complexes. On the whole, the calculation shows good agreement between BONUS and alternative codes; the standard deviation variance is 6–13.5% for fission product mass and activity, which is comparable to the discrepancies between different variants of the alternative calculations themselves.     

Lumped pseudo fission products during burnup step in MCNP5-ORIGEN coupling system

Depletion calculation and accurate inventory of fission products in a nuclear system are required for criticality, safety and spent fuel management. Actual trend is to use Monte Carlo methods. It is well known that the fission process produces a large number of nuclides, some of which have a significant impact on the nuclear properties of the core and its behavior. In this study, we propose to determine the influence of fission products on the behavior of the IAEA 10 MW benchmark reactor. Even if nowadays we have powerful computing capability and we can solve the full system of fission products, such calculations are cumbersome and not needed because most of fission products have low absorption rates and therefore their precise concentrations calculation are not required. The practice is to identify and use only the nuclides which can have a significant absorption cross section.From the entire fission products of the available fissionable actinides, 214 nuclides have been considered. Their selection was essentially based on their absorption rates. To carry out the calculation, 81 were treated explicitly and 133 were lumped into pseudo fission products.A computational method has been developed for burnup and criticality calculations using MCNP5-ORIGEN coupling scheme. The MIXE_ACE program was developed and incorporated within this coupling scheme in order to mix and rewrite in ACE format the selected cross sections of the pseudo fission products for each burnup step. The mass weight of the constituent nuclides was used. The initial one group cross sections library for ORIGEN was generated using average flux spectrum in the core.Using the above methodology, an estimation of k_eff and cross sections during depletion calculations has been carried out for the IAEA 10 MW reactor based on UZrH1.6 fuel. The results are compared to those of ANL (Argonne National Laboratory), MCNP6 and other calculations by using selected fission products from WIMS library. Generally, the results are satisfactory but some discrepancies exist. The differences can be explained mainly by the nature of the fission products considered in the calculation and especially their cross sections.     

Taking burnup credit into account in criticality studies: the situation as it is now and the prospect for the future

As enrichment of the fuel has become higher than the limits used at the designing stages, it seemed necessary to consider fuel depletion during irradiation to guarantee the criticality safety for relatively high enriched fuels transportation, storage or reprocessing. This burnup credit will make it possible to use the devices for spent fuels which are initially relatively high enriched. For that purpose, a method was developed considering: (i) partial Uranium-and-Plutonium burnup credit in the criticality studies, and (ii) a conservative assumption concerning the axial profile; this actinides-only method was supported by an experimental program called HTC. The method was accepted by the French Safety Authority. Moreover, in order to reduce again the calculated values of the reactivity for irradiated fuels, a French working group was set up in 1997 to define a conservative method which enables industrial companies to take burnup credit into account with some of the fission products and using a more precise profile. The work of this group has been divided into four tasks related to: the determination of (i) the composition of the fuel, (ii) a conservative profile, (iii) a conservative irradiation history, and (iv) the calculation scheme. This work is also supported by experimental programs related to the validation of the fission products effects, in terms of reactivity.     

Study of a self-regulated nuclear burn wave regime in a fast reactor based on a thorium–uranium cycle

The possibility of a wave of slow nuclear burning in a fast reactor in thorium–uranium fuel cycle is investigated. The calculations were performed using a model based on the solution of a nonstationary nonlinear diffusion equation for a cylindrical homogeneous reactor using the concept of a radial geometric factor (buckling) and the effective multigroup approximation taking account of the nuclear kinetics of the precursors of delay neutrons and burnup and production of the main nuclides of the thorium–uranium fuel cycle. The calculations showed that the generation and propagation of a wave of nuclear burning traveling with velocity approximately 2 cm/yr are possible in a thorium–uranium medium. However, the addition of even small quantities of a construction material and coolant to the composition of the reactor makes it impossible to obtain the burn wave regime. A self-maintained nuclear burn regime is also established in this case and exists for a long time (∼5 yr), but the system does not transition into a regime with a nuclear burn wave propagating along the axis of the reactor.     

Determination of the Fuel-Assembly Burnup in a Research Reactor by Repeated Short-Time Irradiation Followed by γ Spectrometric Measurements

Experiments performed to determine the absolute fuel burnup in spent fuel assemblies in the IRT research reactor at the Moscow Engineering Physics Institute are described. The method is based on measuring the residual amount of ²³⁵U in the spent fuel asemblies with respect to the activity of the fission product ¹⁴⁰La accumulated in fresh and spent fuel assemblies after they were irradiated for a short time in the reactor core. A fresh fuel assembly with known uranium mass was used as a standard. The neutron flux was monitored using Al + Cu and Al + Co wires placed at the center of the fuel assembly. Small corrections for the difference in the spectrum amd the flux density of the neutrons in fuel assemblies with different uranium content were obtained from the calculations. The burnup of the three fuel assemblies studied was determined to within less than 2%.     

基于抽样方法的燃耗计算核素积存量不确定度分析

基于抽样基本原理研究了应用于燃耗计算的不确定度分析方法,并开发了燃耗计算不确定度分析程序。基于评价核数据库ENDF/B-Ⅷ.0的裂变产额标准差和衰变常量标准差计算得到了衰变常量协方差矩阵和带相关性的裂变产额协方差矩阵,并结合SCALE6.2程序包的56群反应截面协方差数据库,对Takahama-3压水堆组件基准题中SF95-4样品进行不确定度分析。计算了反应截面、衰变常量和裂变产额不确定度引起的核素积存量的不确定度。计算结果表明,反应截面的不确定度是锕系核素积存量不确定度的主要来源,裂变产额和衰变常量的不确定度对部分裂变产物的积存量会引入较大的不确定度。但考虑裂变产额相关性后,裂变产额引起的不确定度显著降低。     

Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

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.