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.     

Effect of Pu-Rich Agglomerate in MOX Fuel on a Lattice Calculation

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 17 × 17 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.     

超临界水冷堆MOX燃料特性分析

针对超临界水冷堆组件,采用不同Pu含量的MOX燃料进行组件计算,得到不同燃料条件下的燃耗深度、功率分布因子、慢化剂温度反应性系数等结果,并对比分析在超临界水冷堆中应用MOX燃料与应用UO2燃料对组件性能的影响,以及不同Pu含量MOX燃料间的性能区别。分析结果表明,在超临界水冷堆设计中,应用MOX燃料与应用UO2燃料有相似的功率分布,应用MOX燃料可以增加燃耗深度,并有良好的慢化剂温度反应性系数。经过合理设计的MOX燃料可较好应用于超临界水冷堆中,且产生更好的性能。     

球床式高温气冷堆MOX燃料循环优化

与压水堆相比,球床式高温气冷堆能在堆芯结构不做明显改变的情况下采用全堆芯装载混合氧化物（MOX）燃料元件。基于250 MW球床模块式高温气冷堆堆芯结构,设计了4种球床式高温气冷堆下MOX燃料循环方式,包括铀钚混合的燃料球和独立的钚球与铀球混合装载的等效方式,采用高温气冷堆设计程序VSOP进行分析,比较了初装堆的有效增殖因数、燃料元件在堆芯内滞留时间、卸料燃耗、温度系数等主要物理特性。结果表明：采用纯铀和纯钚两种分离燃料球且铀燃料球循环时间更长的方案,平均卸料燃耗较高,总体性能较其他循环方式优越。     

MOX燃料堆芯热工特性及设计限值研究

使用MOX燃料的快堆核电站以其线功率高、燃耗高、堆芯出口温度高等特点,对堆芯热工设计提出了新的问题.本文在对MOX燃料热工性能分析的基础上,给出了主要的热工设计限值,并以电功率870 MW电站为参考,初步分析了其堆芯热工特性和设计裕量.结果表明对于MOX燃料,较高的堆芯热工参数合理可行,且具有足够的裕量.     

Analysis of core physics experiments on irradiated BWR MOX fuel in the REBUS program

Critical experiments were performed in the REBUS program on a core loaded with a test bundle including 16 irradiated BWR-type MOX rods of average burnup of 61 GWd/t. The experimental data were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 or JENDL-3.3. Biases in effective multiplication factors of the critical cores were ?1.0%Δk for the diffusion calculations (JENDL-3.2), ?0.3%Δk for the transport calculations (JENDL-3.3), and 0.2%Δk for the Monte Carlo calculations (JENDL-3.2). The measured core fission rate and co-activation rate distributions were generally well reproduced using the three types of calculations. The burnup reactivity determined using the measured water level reactivity coefficients was ?2.41 ± 0.08%Δk/kk’, which also agreed with the results of the three type of calculations within the measurement and calculation errors. The most probable isotopic inventories in the irradiated MOX rods was tentatively obtained by using the ratios of the calculation to chemical assay data on a pellet sample, and the burnup reactivity was reanalyzed to split the calculation error into those due to the inventory and reactivity calculations. This approach showed that the inventory calculation error compensated the reactivity calculation error.     

Estimation of Spent Fuel Compositions from Light Water Reactors

The compositions and quantities of minor actinide (MA) and fission product (FP) in spent fuels will be diversified with the use of high discharged burnup fuels and MOX fuels in LWRs which will be a main part of power reactors in future.

In order to investigate above diversities, we have studied on the calculation method to be used in the estimation of spent fuel compositions and adopted the real irradiation calculation in which axial burnup and moderator distribution are considered in the burnup calculation.

On the basis of the calculations, compositions and burnup quantities of various LWR spent fuels (reactor type: PWR and BWR, discharged burnup: 33, 45 and 60 GWd/tHM, fuel type: U0₂ and MOX) are apparently estimated among various forms of fuels. As an example, it is shown that there are considerable discrepancy in MA burnup between PWR and BWR spent fuels.     

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.     

Statistical Uncertainty Analysis Applied to Fuel Depletion Calculations

In this paper, we present an uncertainty methodology based on a statistical approach, for assessing uncertainties in lattice code predictions of fuel composition changes with burnup due to uncertainties in the fuel geometric configuration, initial enrichment, and depletion conditions. The methodology has been applied to depletion calculations with CASMO-4 and experimental data from the ARIANE Programme to estimate the calculation uncertainties in nuclide concentration and other neutronic parameters at any time during the irradiation history. Results have shown that important information on the quality of the code's predictions can be obtained by analyzing the comparison of the code's estimates and their associated uncertainties, in the form of tolerance intervals, with experimental data and their reported errors.     

MOX燃料组件装入现役M310堆芯对堆芯核设计的影响研究

国际上的MOX燃料技术目前已较为成熟,且已有在压水堆中运行的工程经验。本文对MOX燃料组件的中子学性能进行了分析,对其在我国现役M310堆芯应用的可行性进行了研究,得到了M310堆芯由全部使用UO2燃料组件向使用30%的MOX燃料组件过渡的堆芯燃料管理方案,并对使用MOX燃料组件的堆芯的部分中子学参数进行了初步分析。结果表明:使用30%的MOX燃料组件的堆芯可达到与全UO2堆芯相当的循环长度;堆芯反应性控制能力可满足要求;慢化剂温度系数、Doppler温度系数、Doppler功率系数、氙和钐的动态特性均趋向使堆芯运行更加安全和稳定。本文的研究结果可为MOX燃料在M310堆芯中应用的进一步研究提供参考。     

Sensitivity Study of Fuel Cost in Extended Burnup BWR Core

A sensitivity study on the fuel cost of an extended burnup BWR core has been carried out on the basis of a realistic model of discharge burnup extension. Full power operating length in months in a refueling cycle and the number of refueling batches are chosen as independent variables in the model to describe extended burnup cores of various types. The reference core for the sensitivity study adopts 9-month full power operation and 4-batch refueling scheme. The difference in the plant cost between the extended burnup core and the reference core, which is referred to as plant capacity factor (PCF) credit, is estimated and combined with the fuel cost to calculate the fuel cost with PCF credit.

The results show that the fuel cost with PCF credit decreases for the extended burnup core with stretched operating length as the burnup extends in cases of constant non-operating length in a cycle, and that it may increase for the extended burnup core with decreased batch number in cases of constant plant capacity factor. It is also suggested that the cost minimum combination of the independent variables can be found to a given discharge burnup for the extended burnup core with decreased batch number in an intermediate case between these two extreme cases. Extended burnup cores with fixed batch number tend to have a lower natural uranium requirement, but larger separative work requirement.

The economic break-even condition for the extended burnup core with decreased batch number is discussed based on the fraction of fixed part in the non-operating length, which is insensitive to the cycle length stretch-out.     

Fission Product Model for BWR Analysis with Improved Accuracy in High Burnup

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 0.1Δk 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.     

混合氧化物燃料应用前景及核燃料循环战略

本文叙述了不同国家在不同核电发展条件下的核燃料循环战略初步考虑,并着重叙述了在快堆商业化进程推迟后所发展的混合氧化物(MOX)燃料的设计、制造技术及定量经济分析的基本情况,并建议我国近期应主要采用中间贮存并配以建造后处理中间试验工厂,为今后的商业后处理厂做好技术准备,同时还要跟踪 MOX 燃料技术。     

A study of the generation of 232U in UO2 and MOX fuels

To clarify the generation pathway of ²³²U, an important nuclide for dose evaluation at various stages in the reuse of uranium, concentrations of ²³²U generated through various pathways were evaluated for UO₂ and mixed oxide (MOX) fuels. Burnup calculation was conducted with ORIGEN2.2 code adopting ORLIBJ40 library, a set of cross-section libraries based on JENDL-4.0. It was found that differences in ²³²U concentrations in UO₂ and MOX fuels mainly arise from differences in the initial compositions of ²³⁴U, ²³⁵U, and ²³⁶U. It was also found that the contribution of plutonium and americium isotopes in MOX fuels is small compared with that of uranium isotopes. The results clarified that the capture cross sections of ²³⁰Th, ²³¹Pa, ²³⁵U, and ²³⁶U, as well as the (n,2n) cross sections of ²³⁷Np and ²³⁸U, have a large effect on the generation of ²³²U. Additional investigation showed that ²³²U concentration is strongly affected not only by time after irradiation but also by time before irradiation.     

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

The temperature measurements of mixed oxide (MOX) and UO₂ fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO₂ fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel.     

A Numerical Analysis for a BWR Fuel Assembly Drop Event

A BWR fuel assembly dropped from the crane hook during outage and clashed against the rack bottom plate of spent fuel pool. The area monitoring system indicated no radiation release, however, damage at the top of fuel channel was found in the following inspection. As fuel integrity is essential for further management, a finite element model was established to evaluate the damage condition. Several component elements including fuel rods, tie plates, and channel were set up and integrated into a full assembly. The analysis results provided the impact force on the fuel assembly and the dynamic response of each component element. The event did result in the damage of fuel channel yet fuel rods fracture was not expected. It's consistent with the inspection observations. There is therefore no concern for future operation such as interim dry storage.     

Irradiated Fuel Behavior under Power Oscillation Conditions

In order to examine high burnup fuel performance under power oscillation conditions, two tests of irradiated fuels under simulated power oscillation conditions were conducted in the Nuclear Safety Research Reactor (NSRR). Irradiated fuels at burnups of 56 and 25 GWd/tU were subjected to four to seven power oscillations, which peaked at 50 to 95 kW/m with intervals of 2 s. The power oscillations were caused by quick withdrawal and insertion of six regulating rods of the NSRR with a computerized control. Impacts of cyclic loads on the fuel performance under hypothetical unstable power oscillations arising during an anticipated transient without scram (ATWS) in boiling water reactors (BWRs) were examined in the tests. Deformation of the fuel cladding of the test rods was comparable to those observed in shorter transient tests, which simulated reactivity-initiated accidents (RIAs), at the same fuel enthalpy level up to 368 J/g. The fuel deformation was mainly caused by pellet-cladding mechanical interaction (PCMI) and was roughly proportional to the fuel enthalpy. Enhanced cladding deformation due to ratcheting by the cyclic load was not observed. Fission gas release, on the other hand, was considerably smaller than in the RIA tests, suggesting different release mechanisms in the two types of transients.     

MOX燃料贮存水池核临界安全分析

利用MONK程序对MOX热室项目燃料贮存水池进行了核临界安全分析。针对给定的水池尺寸和燃料棒数量,确定燃料以分区方式贮存。选取国际公布的临界基准实验数据,验证并确定MONK程序计算分析类似物料形态时的偏倚和次临界限值,其次进行保守假设,确定贮存水池在正常及事故工况下其中子有效增殖因数,评价贮存水池的安全性。计算结果表明,贮存水池在最危险事故工况下,其最大中子增殖因数小于次临界限值,系统处于临界安全状态。     

Analysis of MOX Fuel Behavior in Halden Reactor by FEMAXI-6 Code

Capabilities of the FEMAXI-6 code to analyze the behavior of high burnup MOX fuels in LWRs have been evaluated. Coolant conditions, detailed power histories and specifications of the MIMAS-MOX fuel rods, rod 10 and rod 11, of IFA-597.4–7 irradiated in the Halden reactor were input, and calculated rod internal pressures and pellet center temperatures were compared with the measured data for the range of 0-31 MWd/kgUO₂. Some sensitivity studies were conducted mainly with respect to pellet thermal conductivity and swelling rate to investigate the changes in thermal behavior and their effects on fission gas release.

In the irradiation period up to about 23 MWd/kgUO₂, the calculated pellet center temperatures sufficiently agreed with the measured data and also the calculated rod internal pressures reproduced the tendency of an increase in the measured rod internal pressures. These results suggest that fission gas release from MOX fuels can be reasonably predicted by a diffusion process that is modeled in UO₂ pellet grains. On the other hand, the steep increase in the measured rod internal pressures observed at the power ramp around 23 MWd/kgUO₂ cannot be reproduced by FEMAXI-6 and can be regarded as the result of a relatively large amount of gas release, which possibly caused a pellet-cladding-gap closure through pellet gas-bubble swelling.     

Analysis of Core Physics Experiments on Fresh and Irradiated PWR UO2 Fuels in the REBUS Program

Critical experiments of two cores each loaded with fresh 5 × 5 test PWR-type fuel rods of ²³⁵U enrichment of 3.8 wt% or irradiated 5 × 5 test rods of rod average burnup of 55 GWd/t in the REBUS program were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 and JENDL-3.3. Biases in effective multiplication factors k _eff's of the critical cores were about ?1:2%Δk for the diffusion calculations (JENDL-3.2), ?0:5%Δk for the transport calculations (JENDL-3.3), and ?0:5 and 0.1%Δk for the Monte Carlo calculations (JENDL-3.3 and JENDL-3.2, respectively). The measured core fission rate and Sc- or Co-activation rate distributions were generally well reproduced using the three types of calculation. The burnup reactivity determined using the measured water level reactivity coefficients was ?2:35 ± 0:07Δk/kk′. The calculated result of the Monte Carlo calculations agreed with it; however, the diffusion and transport calculations overestimated the absolute value by about 7%, which would be mainly attributed to the errors in the calculation of the reactivity caused by changing the fuel compositions from fresh fuel to irradiated fuel.