Analysis of fuel temperature effects on reactivity of light water reactor fuel assemblies by using MVP-2 adopting an exact resonance elastic scattering model

Using the continuous-energy Monte Carlo code MVP-2 adopting a resonance elastic scattering model considering the thermal motion of a target nucleus (the exact model) for major heavy nuclides, analysis of fuel temperature effects on reactivity of mockup UO₂ and MOX fuel assemblies for light water reactors was performed, and the results were compared with those of the conventional asymptotic model. A base condition was a hot operating condition with an in-channel void fraction of 40% and fuel temperature of 520 ℃ for the BWR fuel assemblies and a hot zero-power condition with fuel temperature of 284 ℃ for the PWR fuel assemblies. The fuel temperature of a high-temperature condition was 1500 ℃ for both types of assemblies. The calculated results showed that the exact model made the neutron multiplication factors at the high-temperature condition lower by ?220 to ?440 pcm (10^?5 Δk) and the Doppler reactivity between the base- and high-temperature conditions more negative by 7% to 10% compared with those obtained by the asymptotic model. The energy-dependent reaction rates of capture and ν-fission were also analyzed to study the detail mechanism in the effect of the exact model on the assembly reactivity.     

Interpretation of the VENUS VIP-BWR program with APOLLO2.8 reference and optimized MOC schemes for 8 × 8 UO2 and MOX BWR assemblies

The interpretation of the VIP-BWR program conducted in the CEN·SCK Mol VENUS critical facility (Belgium), has been performed with the new APOLLO2.8 product and its CEA2005V4.1 library based on the JEFF3.1.1 file. Both reference SHEM-MOC (281groups without equivalence) and Optimized BWR 26G (26 groups with equivalence) schemes are used for UO₂ and MOX BWR assembly calculations. The VIP-BWR program was aimed to provide an experimental database for BWR neutronics tools in mixed Gd poisoned configurations with 8 × 8 UO₂ and MOX assemblies. The experimental conditions are relatively representative of actual industrial BWR core characteristics, at least in terms of void fraction. Measured pin-by-pin power distributions enable to exact valuable information at various interfaces. For fresh (UO₂/UO₂–Gd) and recycled UO₂ (UO₂ only) cores loadings, the information is given through the “UO₂” core. In the case of partial MOX loadings (UO₂/MOX interface), the power distributions are available through the “T-MOX” core. All critical sizes are predicted within 1 with SHEM-MOC reference calculation scheme. For UO₂ core, the (C–E) on k_eff are (95 ± 266) pcm and (203 ± 266) pcm for SHEM-MOC and Optimized scheme respectively. For MOX core, the results are (87 ± 214) pcm and (283 ± 214) pcm. The uncertainties take into account both measurement uncertainties and technological uncertainties such as enrichment, clad thicknesses, grid pitch or fuel densities.     

Analysis of measured isotopic compositions by CASMO5 coupled with a JENDL-4.0 base library for irradiated fuel of light water reactors

Measured isotopic compositions of UO₂ and MOX fuel samples taken from irradiated light water reactor fuel assemblies were analyzed by CASMO5 coupled with a JENDL-4.0 base library to assess the uncertainties in the calculated isotopic compositions on heavy and fission product nuclides. The burnup calculations for the analysis were performed based on a single-assembly model taking into account the detail fuel assembly specifications and irradiation histories. For the MOX fuel samples, a multiple-assembly model was also adopted taking into account the effect of the surrounding UO₂ fuel assemblies. The average and standard deviation of the biases (C/E ? 1's (here C and E are calculated and measured results, respectively)) were calculated for each nuclide separately on the PWR and BWR UO₂ fuel samples. The averaged biases for ²³⁵U, ²³⁶U, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu and ²⁴²Pu were 2.7%, ?0.9%, 0.3%, 0.7%, ?2.4% and ?1.7% for PWR UO₂ samples, and 6.7%, ?1.5%, 2.5%, ?0.6%, 0.4% and ?0.1% for BWR UO₂ samples, respectively. The biases with the single-assembly model on the MOX fuel samples showed large positive values of ²³⁹Pu, and application of the multiple-assembly model reduced the biases as reported in our previous studies.     

Performance improvement analysis of boiling water reactors by incorporation of hydride fuel

The feasibility of improving the neutronic characteristics of boiling water reactors (BWR) by using U–Zr hydride fuel is studied. Several modified BWR fuel assembly designs are considered. These include designs in which hydride fuel rods replace water rods only, replace water rods and a fraction of the oxide fuel rods, replace oxide fuel in the upper half of all the fuel rods, and replace all the oxide fuel in the assembly. It is found that replacement of at least half of the oxide fuel rods in the fuel assembly by U–ZrH_1.6 fuel might simultaneously improve the performance of BWR in three ways: (a) Increasing the energy extracted per fuel assembly and the cycle length by up to 10%. (b) Reducing the uranium ore and SWU requirements by approximately 10%. (c) Reducing the negative void coefficient of reactivity by, at least, 50%. It is also found that replacement of all the oxide fuel by hydride fuel opens interesting new options for the design of BWR fuel assemblies. The net result might be simplified assembly designs that can generate significantly more energy while featuring small negative void coefficient of reactivity. U–ThH₂ fuel appears to be even more promising than U–ZrH_1.6. For the potential benefits from hydride fuel to be realized, a clad material that is not permeable to hydrogen and is not as neutron absorbing as stainless steel needs to be developed.     

Analysis of Measured Isotopic Compositions of High-Burnup BWR MOX Fuel

Analysis of measured isotopic compositions of four high-burnup BWR MOX fuel samples was performed by using a general-purpose neutronic calculation code SRAC and a continuous-energy Monte Carlo burnup code MVP-BURN. The initial Pu fissile content of the samples was 5.52 wt%, and the burnups ranged from 50 to 80 GWd/t. It is confirmed that a geometrical model including the effect of UO₂ assemblies adjacent to the MOX assembly is necessary in the burnup calculations to obtain accurate calculated isotopic compositions. The calculated results of MVP-BURN with JENDL-3.3 taking such effect into account show more accurate results for major actinides (U, Pu, and Am isotopes) and most fission products than those of infinite assembly calculations. The paper also shows the results calculated using SRAC with JENDL-3.3, ENDF/B-VII, and JEFF-3.1.     

Design of an overmoderated fuel and a full MOX core for plutonium consumption in boiling water reactors

The use of uranium–plutonium mixed oxide fuel (MOX) in light water reactors (LWR) is nowadays a current practice in several countries. Generally 1/3 of the reactor core is loaded with MOX fuel assemblies and the other 2/3 with uranium assemblies. Nevertheless the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this paper the design of a boiling water reactor (BWR) core fully loaded with an overmoderated MOX fuel design is investigated. The design of overmoderated BWR MOX fuel assemblies based on a 10×10 lattice are developed, these designs improve the neutron spectrum and the plutonium consumption rate, compared with standard MOX assemblies. In order to increase the moderator to fuel ratio two approaches are followed: in the first approach, 8 or 12 fuel rods are replaced by water rods in the 10×10 lattice; in the second approach, an 11×11 lattice with 24 water rods is designed with an active fuel length very close to the standard MOX assembly. The results of the depletion behavior and the main steady state core parameters are presented. The feasibility of a full core loaded with the 11×11 overmoderated MOX fuel assembly is verified. This design take advantage of the softer spectrum comparable to the 10×10 lattice with 12 water rods but with thermal limits comparable to the standard MOX fuel assembly.     

Analysis of Mixed Oxide Fuel Critical Experiments with Neutronics ANalysis Codes for Boiling Water Reactors

Critical experiments of UO₂ and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were aNalyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library.

The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%ΔAk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO₂ and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO₂ cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT.

These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO₂ cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP.     

Zircaloy-4 cladding corrosion model covering a wide range of PWR experiences

A phenomenological corrosion model for Zircaloy-4 cladding was developed by focusing on the effect of the metallurgy of cladding and the water chemistry combined with the thermo-hydraulic conditions. The metallurgical effect was formulated by considering the Sn content in the cladding and the heat treatment of the cladding. Concerning the effect of the water chemistry, it is assumed that lithium and boron have an influence on the corrosion under the condition of subcooled void formation on the cladding surface. The developed corrosion model was implemented in a fuel performance code, COSMOS, and verified using the database obtained for the UO₂ and MOX fuel rods irradiated in various PWRs. It was elucidated that the corrosion by lithium was enhanced in the case where the fuel rods were irradiated with a high linear power so that a significant subcooled void could be formed on the cladding surface. On the other hand, there was no evidence of the lithium effect even though its concentration was high enough if the void in the coolant was negligible. This result shows that the acceleration of corrosion by an increased lithium concentration occurs only when subcooled voids are formed on the cladding surface. In addition, the comparison between the measurement and the prediction for the MOX fuel rods indicates that no distinguishable difference is found in the corrosion behavior between the MOX and the UO₂ fuels as expected.     

基于SCIENCE程序包的IFBA组件模型的可行性研究

表面涂有一薄层硼化锆的一体化燃料可燃吸收体(IFBA)被用作轻水堆UO2燃料组件的反应性控制。法国AREVA公司开发的SCIENCE程序包具有模拟IFBA组件的能力,但其模拟精度需经标定。本文利用APOLLO2-F程序建立IFBA组件模型和不含IFBA组件模型,研究了组件的无限增殖因数k∞及IFBA价值,并与西屋公司结果进行比较。分析了燃料和包壳温度的处理方法以及数据库的差异对结果的影响。利用硼化锆密度修正因子评估IFBA价值偏差对堆芯参数和功率分布等的影响。结果表明:SCIENCE计算的k∞及IFBA价值与西屋公司的结果符合较好,低燃耗区SCIENCE计算的价值偏小2%。装载8个104根IFBA棒组件的堆芯,组件相对功率最大偏差约为1%;硼浓度、功率峰因子FQ和焓升因子FΔH的变化均不到0.1%,可忽略。先导组件采用28根或更少的IFBA棒时,可直接采用SCIENCE程序进行计算。     

Monte Carlo Modelling of Increasing Void Fraction in 100% MOX ABWR: Lessons Drawn from the FUBILA Program

The French Atomic Energy Commission CEA and the Japanese Incorporated Administrative Agency JNES (Japan Nuclear Energy Safety Organisation) have undertaken first-of-a-kind full MOX core physics experiments, FUBILA, in the EOLE critical facility of the CEA Cadarache Centre. The experiments have been designed to obtain core physics data under high-burn-up 9 × 9 and 10 × 10 BWR MOX assemblies operating conditions. The experimental program, consisting of eight different core configurations, started in January 2005 and ended on September 1, 2006. The analysis of the void increase part of the experimental data between 0 and 70% void has been carried out using the French TRIPOLI-4.5 continuous-energy Monte Carlo calculation code with the newly released JEFF3.1.1 nuclear data library. The average C/E discrepancies obtained on critical masses, reactivity worth, and pin-by-pin power distributions enable us to estimate all the integral and local parameters with uncertainties largely within the target uncertainties, demonstrating the capability of the code to treat complex geometries with a high degree of accuracy. Additional k_eff calculations performed with the latest ENDF/B-VII evaluation exhibit a clear tendency to overestimate the k_eff by about 500 to 650 pcm and the void worth by more than 4%, showing that the JEFF3.1.1 library is more precise for MOX lattices.     

AP1000 core design with 50% MOX loading

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO₂ fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO₂ core design and a mixed MOX/UO₂ core design, discussing relevant results related to reactivity management, power margin and fuel rod performance.     

An Apparatus for Continuously Measuring Radon-222 Exhalation from Ground

High burnup MOX and UO₂ test rods were prepared from the fuel rods irradiated in commercial BWRs. Each test rod was equipped with a fuel center thermocouple and reirradiated in the Halden boiling water reactor (HBWR) in Norway. The burnups of MOX and UO₂ test rods reached about 84GWd/tHM and 72GWd/t, respectively. Fuel temperature was measured continuously during the re-irradiation tests. Thermal conductivity change in high burnup fuel was evaluated from the results of comparison between the measured fuel temperature and the data calculated by using the fuel analysis code FEMAXI-6. The comparison results suggested that the thermal conductivity of MOX fuel pellets is comparable to that of UO₂ fuel pellets in the high burnup region around 80 GWd/t. It is probable that the impurity effect of Pu atoms gradually diminishes with increasing burnup because other factors that affect pellet thermal conductivity, such as the accumulation effect of soluble fission products and irradiation-induced defects in crystal lattice, become dominant in a high burnup region.     

Analysis of measured isotopic compositions of high-burnup PWR MOX and UO2 fuels in the MALIBU program

The measured isotopic compositions of fuel samples taken from high-burnup spent PWR MOX and UO₂ assemblies in the MALIBU program has been analyzed by lattice physics codes. The measured isotopes were U, Np, Pu, Am, and Cm isotopes and about 30 major fission product nuclides. The codes used in the present study were a continuous-energy Monte Carlo burnup calculation code (MVP-BURN) and a deterministic burnup calculation code (SRAC) based on the collision probability method. A two-dimensional multi-assembly geometrical model (2 × 2 model) was mainly adopted in the analysis in order to include the fuel assemblies adjoining the relevant fuel assembly, from which the samples were taken. For the MOX sample, the 2 × 2 model significantly reduces the deviations of the calculated results from the measurements compared with a single assembly model. The calculation results of MVP-BURN in the 2 × 2 model reproduce the measurements of U, Np, and Pu isotopes within 5% for the MOX sample of 67 GWd/t. The deviations of their calculated results of U, Np, and Pu isotopes from the measurements are less than 7% for the UO₂ sample of 72 GWd/t.     

Application of Multiband Method to Pin Cell Calculations

Abstract

In order to accurately calculate effective neutron cross sections in the resonance energy region, the multiband method has been applied to cell calculations. Cell calculations for UO² and MOX fuels of light water reactors have been performed and the results were compared with those of a continuous energy Monte Carlo code VIM and the conventional self-shielding method using the Dancoff factor.

The k∞values calculated by the multiband method agreed with those of the VIM calculations within 0.20% Δk for the UO² fuel cell and within 0.30% Δk for the MOX fuel cell, respectively, whereas the Dancoff factor method yielded about l.l%Δk errors for the two cells. The element- wise contribution to this error was investigated, and it was found that the effective microscopic cross sections, particularly those for the giant resonances of ²³⁸U, calculated by the multiband method were in good agreement with those of VIM. It was also found that interference effect between ²³⁸U and ²³⁵U resonances in the UO² fuel and that between ²³⁸U and ²³⁹Pu resonances in the MOX fuel made about 0.20%Δk contributions to k∞ in both fuel cells.     

Nuclear Analysis of PIE Data of Irradiated BWR 8×8-2 and 8×8-4 UO2 Fuel Assemblies

The measured pellet average inventories of actinides and fission product nuclides on the fifteen samples taken from a three-cycle irradiation BWR 8×8-2 UO₂ assembly were compared with those of assembly burnup calculations using a collision probability method (SRAC) with the JENDL-3.2 nuclear data library. The present calculations overestimate the inventories of ²³⁵U, well reproduce those of ²³⁹Pu and ²⁴⁰Pu, yet underestimate those of ²³⁶U, ²³⁷Nd, ²³⁸Pu, ²⁴¹Pu, and ²⁴²Pu. The inventories of minor actinides are underestimated by the present analysis except for ²⁴¹Am. The major FP nuclides contributing to neutron absorption such as Nd, Cs, Eu, and Sm are almost well reproduced by the present calculations. The measured pellet average burnups and major actinide inventories on the twenty samples taken from four BWR 8×8-4 UO₂ assemblies were also compared with those of the burnup calculations using SRAC and a continuous energy Monte Carlo burnup analysis code (MVP-BURN). Most of the calculated pellet average burnups of both codes agree with the measurements within the range of ±10%. The general trends of the measured pellet radial distributions of actinide and FP nuclides on six samples of the 8×8-4 UO₂ assemblies were well reproduced by the burnup calculations of MVP-BURN.     

Low Current Efficiency in MOX Deposition Tests

A series of MOX deposition tests has been performed since 2001 at RIAR to clarify its complex phenomena and to improve its poor current efficiency. In the 2001 tests, the cathode current efficiency was between 60 and 100% but the Pu fraction in the MOX was between 5 and 20%. In 2002 tests, the fraction was raised to more than 30% by modifying the test conditions but the current efficiency fell to between 20 and 60%. A new method was proposed to simulate the parasitic current due to the electrode reactions of UO₂ ²+/UO₂+, Pu⁴+/Pu³+ and Fe³+/Fe²+ at the cathode. It was found that the parasitic current due to the UO₂ ²+/UO₂+ reaction significantly lowers the current efficiency especially when the cathode potential is kept near the equilibrium value during the electrolysis to increase the Pu fraction in the MOX deposit.     

MOX Fuel Performance Experiment Specified for Japanese PWR Utilisation in the HBWR

In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pufissile enrichment of about 6 wt% have been irradiated in the HBWR. In-pile performance data of MOX have been obtained, and the peak burn-up of MOX pellet have reached to 66 GWd/tM as of October 2004. MOX fuel temperature is confirmed to have no significant difference compared to UO₂, if taking into account adequately for thermal conductivity degradation due to PuO₂ addition and burn-up development, and measured fuel temperature agrees well with HB-FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly larger than UO₂ based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behaviour. MOX fuel swelling rate agrees well with solid swelling rate. Cladding elongation data shows onset of PCMI in high power region. Ramp test data from other experiment programs with various types of MOX fabrication route confirms superior PCI resistance of MOX compared to UO₂, due to enhanced creep rate of MOX. The irradiation is expected to continue until achieving of 70 GWd/tM (MOX pellet peak).     

Benchmark Problem Suite for Reactor Physics Study of LWR Next Generation Fuels

This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO₂ and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management.     

A VVER-1000 LEU and MOX assembly computational benchmark analysis using the lattice burnup code EXCEL

Utilization of Mixed Uranium–Plutonium Oxide (MOX) fuel in VVER-1000 reactors envisages the core physics analysis using computational methods and validation of the related computer codes. Towards this objective, an international experts group has been established at OECD/NEA. The experts group facilitates sharing of existing information on physics parameters and fuel behaviour. Several benchmark exercises have been proposed by them with intent to investigate the core physics behaviour of a VVER-1000 reactor loaded with 2/3rd of low enriched uranium (LEU) fuel assemblies (FA) and 1/3rd of weapons grade mixed oxide (MOX) FA. In the present study an attempt is made to analyse ‘AVVER-1000LEUandMOXAssemblyComputationalBenchmark’ and predict the neutronics behaviour at the lattice level. The lattice burnup code EXCEL, developed at Light Water Reactor Physics Section, BARC is employed for this task. The EXCEL code uses the 172 energy group ‘JEFF31GX’ cross-section library in WIMS-D format. Assembly level fuel depletion calculations are performed up to a burnup of 40 MWD/kg of heavy metal (HM). Studies are made for the parametric variations of fuel and moderator temperatures, coolant density and boron content in the coolant. Both operational and off-normal states are analysed to determine the corresponding infinite neutron multiplication factor (k_∞). Pin wise isotopic compositions are computed as a function of burnup. Isotopic compositions in different annular regions of Uranium–Gadolinium (UGD) pin, fission rate distributions in UGD, UO₂ and MOX pin cells are also computed. The predicted results are compared with the benchmark mean results.     

Failure of high burnup fuels under reactivity-initiated accident conditions

Pulse irradiation experiments of high burnup light-water-reactor fuels were performed to assess the fuel failure limit in a postulated reactivity-initiated accident (RIA). A BWR-UO₂ rod at a burnup of 69 GW d/t failed due to pellet-cladding mechanical interaction (PCMI) in the test LS-1. The fuel enthalpy at which fuel failure occurred was comparable to those for PWR-UO₂ rods of 71 to 77 GW d/t with more corroded cladding. Comparison of cladding metallographs between the BWR and PWR fuel rods suggested that the morphology of hydride precipitation, which depends on the cladding texture, affects the fuel failure limit. The tests BZ-1 and BZ-2 with PWR-MOX rods of 48 and 59 GW d/t, respectively, also resulted in PCMI failure. The fuel enthalpies at failure were consistent with a tendency formed by the previous test results with UO₂ fuel rods, if the failure enthalpy is plotted as a function of the cladding outer oxide thickness. Therefore, the PCMI failure limit under RIA conditions depends on the cladding corrosion states including oxidation and hydride precipitation, and the same failure limit is applicable to UO₂ and MOX fuels below 59 GW d/t.