Application of candle burnup strategy for future nuclear energy utilization

The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top (or from top to bottom) of the core and without any change in their shapes. Therefore, any burnup control mechanisms are not required, and reactor characteristics do not change along burnup. The reactor is simple and safe. If this burnup scheme is applied to some neutron rich fast reactors, either natural or depleted uranium can be utilized as fresh fuel after second core and the burnup of discharged fuel is about 40%. It means about 40% of natural or depleted uranium can be utilized without either enrichment or reprocessing.

In the ideal nuclear energy utilization system, the radioactive toxicity in the environment should remain or decrease after the utilization. This requirement is very severe and difficult to be satisfied. It may take too much time for its realization. The CANDLE burnup may substitute this period. Though it is a once-through fuel cycle, the discharged fuel burnup is about ten times of the present value for light water reactors. The space necessary for final disposal can be drastically reduced. However, in order to realize such a high burnup of discharged fuels some innovative technologies should be developed. Either new material standing still for such a high burnup or intermediate recladding will be required. Especially new fuel development will take a lot of time. For the time being a small reactor with CANDLE burnup may be a good option for nuclear power generation. Even this kind of reactor requires some innovative technologies and a long period for their developments. For the first stage of CANDLE burnup the prismatic fuel high-temperature gas cooled reactor is preferable. Since the design of this reactor fits to the CANDLE burnup very well, only a little time is required for its research and development.     

Reactivity and neutron emission measurements of highly burnt PWR fuel rod samples

Fuel rods with burnup values beyond 50 GWd/t are characterised by relatively large amounts of fission products and a high abundance of major and minor actinides. Of particular interest is the change in the reactivity of the fuel as a function of burnup and the capability of modern codes to predict this change. In addition, the neutron emission from burnt fuel has important implications for the design of transport and storage facilities. Measurements have been made of the reactivity effects and the neutron emission rates of highly burnt uranium oxide and mixed oxide fuel rod samples coming from a pressurised water reactor (PWR). The reactivity measurements have been made in a PWR lattice in the PROTEUS zero-energy reactor moderated in turn with: water, a water and heavy water mixture and water containing boron. A combined transport flask and sample changer was used to insert the 400 mm long burnt fuel rod segments into the reactor. Both control rod compensation and reactor period methods were used to determine the reactivities of the samples. For the range of burnup values investigated, an interesting exponential relationship has been found between the neutron emission rate and the measured reactivity.     

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.     

(JAERI) The First International Conference on Supercomputing in Nuclear Applications

The continuous energy Monte-Carlo/collision probability hybrid method has been developed for efficient burnup calculations of light water reactor fuel assemblies. This hydrid method was applied to the NEACRP LWR fuel burnup benchmark, and the numerical results were in good agreement to those of the reference Monte-Carlo calculations in about 1/5 CPU time compared to the reference one, though there is a large difference between the results of RESPLA⁽¹⁾ (collision probability method) and VIM⁽²⁾ (Monte-Carlo method). Thus this hybrid method is found to be effective for burnup calculations of light water reactor fuel assemblies.     

Long life small CANDLE-HTGRs with thorium

CANDLE (constant axial shape of neutron flux, nuclide densities and power shape during life of energy producing reactor) burnup strategy is applied to small (30 MWth) block-type high temperature gas-cooled reactors (HTGRs) with thorium fuel. The CANDLE burnup is adopted in this study since it has several promising merits such as simple and safe reactor operation, and the ease of designing a long life reactor core. Burnup performances of thorium fuel (²³³U, ²³²Th)O₂ are investigated for a range of enrichment ⩽15%. Discharged fuel burnup and burning region motion velocity are major parameters of its performances in this study. The reactors with thorium fuel show a better burnup performance in terms of higher discharged fuel burnup and slower burning region motion velocity (longer core lifetime) compared to the reactors with uranium fuel.     

Burnup characteristics and fuel cycle economics of mixed uranium–thorium fuel in a simplified small pebble bed reactor

The limitation of natural uranium resources and the improvement of economic values of nuclear reactors are important issues to be solved in the future development of these reactors. In our previous study, we presented an innovative design for simplifying a pebble bed reactor, and the optimization of this design showed that burnup values could be increased and natural uranium uses could be reduced. The purposes of the current study were to design a simplified pebble bed reactor by removing the unloading device from the reactor system and to further optimize the burnup characteristics of this reactor with a peu à peu fuel-loading scheme by introducing thorium in the fuel configuration as a fertile material. Another goal was to optimize the fuel composition so that the system could achieve even better burnup characteristics and use scarce uranium resources more efficiently. Using a specially developed computer code, we analyzed and optimized the performance of a 110-MWt simplified pebble bed reactor using a peu à peu fuel-loading scheme. An optimized design using 30% of fertile thorium mixed with uranium fuel with 15% ²³⁵U enrichment and a 7% packing fraction calculated to achieve a high burnup of 140 GWD/T for more than 21 years' operation time that could save 13 to 33% of natural uranium use compared with the savings noted in our previous study. Neutronic, burnup and fuel economic analysis for this optimized design are discussed in this study.     

径向步进倒料行波堆的数值研究

行波堆是一种可实现自持增殖-燃耗的新概念快堆,它可直接使用天然铀、贫铀、钍等可转换核材料,实现非常高的燃料利用率。基于行波堆的原理,提出了具有现实应用价值的径向步进倒料行波堆的概念,并将其与典型钠冷快堆的设计相结合,采用数值方法对由外而内的径向步进行波堆二维渐近稳态特性进行了研究。计算结果表明:渐近keff随倒料循环周期近似抛物线分布,而渐近燃耗随倒料循环周期线性增长,满足临界条件的倒料循环周期中最大燃耗可达38%;堆芯功率峰随着倒料循环周期的增长,从燃料卸出区(堆芯中心)向燃料导入区(堆芯外围)移动,功率峰值逐渐降低,在高燃耗情况下,靠近堆芯中心的轴向功率分布呈M形。     

The OECD Halden reactor project fuels testing programme: methods, selected results and plans

The fuels testing programme conducted in the Halden reactor (heavy boiling water reactor (HBWR)) is aimed at providing data for a mechanistic understanding of phenomena, which may affect fuel performance and safety parameters. The investigations focus on implications of high burnup and address thermal property changes, fission gas release as influenced by power level and operation mode, fuel swelling, and pellet–clad interaction. Relevant burnup levels (>50 MWd kg⁻¹ U) are provided through long-term irradiation in the HBWR and through utilisation of re-instrumented fuel segments from commercial light water reactors (LWR). Both urania and MOX fuels are being studied regarding thermal behaviour, conductivity degradation, and aspects of fission gas release. Experiments are also conducted to assess the cladding creep behaviour at different stress levels and to establish the overpressure below which the combination of fuel swelling and cladding creep does not cause increasing fuel temperatures. Clad elongation measurements provide information on the strain during a power increase, the relaxation behaviour and the extent of a possible ratcheting effect during consecutive start-ups. Investigations foreseen in the programme period 2000–2002 include the behaviour of MOX and Gd-bearing fuel and other variants developed in conjunction with burnup extension programmes. Some LWR-irradiated fuel segments will undergo a burnup increase in the HBWR to exposures not yet achieved in LWRs, while others will be re-instrumented and tested for shorter durations.     

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.     

Artificial neural network modeling for fission gas release in LWR UO2 fuel under RIA conditions

A fission gas release (FGR) model was developed by using an artificial neural network method to predict fission gas release in UO₂ fuel under reactivity initiated accident (RIA) conditions. Based on the test data obtained in the CABRI test reactor and nuclear safety research reactor, the model takes into account the effect of the five parameters: pellet average burnup, peak fuel enthalpy, the ratio of peak fuel enthalpy to pulse width, fission gas release during base-irradiation, and grain size of a fuel pellet. The parametric study of the model, producing a physically reasonable trend of FGR for each parameter, shows that the pellet average burnup and the ratio of peak fuel enthalpy to pulse width are two of the most important parameters. Depending on the combination of input values for the five parameters, the application of the model to a fuel rod under typical RIA conditions of light water reactor produces 1.7-14.0% of FGR for the pellet average burnup ranging from 20 to 70 MW d/kg U.     

燃耗数据库基准检验方法研究

燃耗数据库基准检验方法对于研制高准确度的燃耗数据库至关重要。本文以TAKAHAMA 3压水堆辐照后检验实验中SF95样品的建模为例，研究了建模要素对燃耗计算的影响，确定了燃耗实验建模的方法，开展了燃耗信用制研究感兴趣的锕系和裂变产物核素积存量计算值与实验值的比对。比对结果显示，主锕系核素计算偏差小于2%，大部分次锕系核素偏差小于10%，大部分重要裂变产物核素偏差小于5%。本文还对125Sb积存量随燃耗深度变化规律进行了理论分析，确认了破坏性放化实验测量结果存在缺陷，并进一步获得了125Sb积存量的修正值，使计算偏差从接近170%下降到20%以内。本次研究表明，燃耗数据库基准检验研究不仅需发展适当的燃耗实验建模方法，还需对实验数据进行适当的评价。     

Re-evaluation of Assay Data of Spent Nuclear Fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems

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.     

Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied.From the result of the burnup calculation, it has been seen that ratio of 40–50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara).By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mm×2, internal blanket of 150 mm and axial blanket of 400 mm×2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internalblanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation.It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mm×2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature coefficient is negative for both of cases.     

Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

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) O₂ and irradiated UO₂ 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.     

Nuclide analysis in high burnup fuel samples irradiated in Vandellós 2

In the framework of a high burnup fuel demonstration programme, rods with an enrichment of 4.5% ²³⁵U were operated to a rod average burnup of about 70 MWd/kgU in the Spanish Vandellós 2 pressurised water reactor. The rods were sent to hot cells and used for different research projects. This paper describes the isotopic composition measurements performed on samples of those rods, and the analysis of the measurement results based on comparison against calculated values.The fraction and composition of fission gases released to the rod free volume was determined for two of the rods. About 8% of Kr and Xe produced by fission were released. From the isotopic composition of the gases, it could be concluded that the gases were not preferentially released from the peripheral part of the fuel column.Local burnup and isotopic content of gamma emitting nuclides were determined by quantitatively evaluating axial gamma scans of the full rods. Nine samples were cut at different axial levels from three of the rods and analysed in two campaigns. More than 50 isotopes of 16 different elements were assessed, most of them by Inductively Coupled Plasma Mass Spectrometry after separation with High Performance Liquid Chromatography. In general, these over 400 data points gave a consistent picture of the isotopic content of irradiated fuel as a function of burnup. Only in a few cases, the analysis provided unexpected results that seem to be wrong, in most cases due to unidentified reasons. Sample burnup analysis was performed by comparing experimental isotopic abundances of uranium and plutonium composition as well as neodymium isotopic concentrations with corresponding CASMO based data. The results were in agreement with values derived independently from gamma scanning and from core design data and plant operating records.Measured isotope abundances were finally assessed using the industry standard SAS2H sequence of the SCALE code system. This exercise showed good agreement between measured and calculated values for most of the analysed isotopes, similar to those reported previously for lower burnup ranges. Thus, it could be concluded, that SAS2H results for high burnup samples are not subject to higher uncertainty and/or different biases than for lower burnup samples, and that the different isotopic experimental measurement methods provide accurate results with acceptable precision.     

Multiobjective Optimization for Expansion Planning of Rad-Waste Management System

In order to investigate fuel behavior under high burnup irradiation condition of high temperature gas-cooled reactor (HTGR), an irradiation test was performed. An irradiation was carried out as a part of a cooperative effort between the US DOE and the Japan Atomic Energy Research Institute. The fuel for the irradiation test was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR). In order to keep fuel integrity up to high burnup over 5%FIMA (% fission per initial metallic atom), thickness of buffer and SiC layers of fuel particle were increased. This report describes the fuel behavior under high burnup condition in the irradiation test.     

商用压水堆核燃料研发进展与应用展望

压水堆(PWR)是目前核电厂反应堆的主力堆型,而核燃料是反应堆的能量源泉和放射性裂变物质的主要来源,关乎核电厂的经济性和安全性。本文对当前国际上面向商用PWR应用研发的掺杂UO₂燃料、高铀密度燃料、微封装燃料和金属燃料的性能特点、技术状态及前景进行了归纳和评价。在掺杂UO₂燃料中,大晶粒燃料具有较高的技术成熟度,将在PWR实现大规模商用;高铀密度燃料和金属燃料在高温水腐蚀氧化问题以及事故下的行为仍待研究解决;具有极致安全的微封装燃料更适合特殊用途的小型反应堆。应协同开展先进燃料组件设计、建立设计准则以及研发高保真的性能分析技术等,以充分发挥新型燃料的可靠性及高燃耗优势。     

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。     

Burnup characteristics and performance of boron nitride and boron coated urania and urania-gadolinia fuels

In this research the burnup performance of (1) boron nitride (BN) and (2) boron nitride-boron (B), hybrid coated urania (UO₂) and urania-gadolinia (Gd₂O₃) fuels were studied. The behavior of fuel burnup, depletion of BN and B, the effect of coating thickness and also Gd₂O₃ content on the burnup performances of the fuels were found by using the code WIMS-D/5 for pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. The optimum thickness ratio of B to BN was found as 4 and their thicknesses were chosen as 4 and 1 μm, respectively. Based on this observation, calculations were performed for the standard fuel assemblies with and without burnable poison and similar assemblies with BN-B hybrid coating with the desired BN-B thicknesses. Results are discussed to make an assessment of the effect of such hybrid coating on fuel cycle characteristics.     

Improvement of power coefficient by using burnable poison in the CANDU reactor

For stable operation of a power reactor, the power coefficient (PC) of the nuclear reactor should be less than or equal to zero. In the CANDU reactor loaded with the recovered uranium (RU) which has a uranium enrichment of ∼0.9 wt% U²³⁵, the PC is estimated to be clearly positive over a wide power range of interest, owing to the generic positive coolant temperature coefficient (CTC) and weak fuel temperature coefficient (FTC) in the CANDU reactor. In order to improve the PC of the CANDU reactor without seriously compromising the economy, introduction of the burnable poison (BP) has been proposed in this work and a physics study has been performed to find the optimal BP material and its optimal loading scheme for the CANDU reactor loaded with the CANFLEX-RU fuel. Four potential BPs (Dy, Er, Eu, and Hf) were evaluated to find the optimal BP and various loading options of the selected BP were evaluated to determine the optimal loading scheme of the BP. From the viewpoint of the achievable fuel discharge burnup, it was found that Er is evaluated to be the best BP and the BP should be loaded within the central two fuel rings because the BP loading on the inner ring is more effective for reducing the CTC. The discharge burnup of the Er-loaded CANFLEX-RU fuel was 38% higher than that of the standard natural uranium (NU) fuel. The fuel discharge burnup can be increased further if the fuel enrichment is increased. It is shown that the discharge burnup of 1.0 wt%-enriched uranium fuel is 1.7 times higher than that of the NU fuel. This study has shown that the use of the BP is feasible to render the PC of the CANDU reactor negative, even though the slight reduction of the fuel burnup is inevitable, and thus the reactor safety can be greatly improved by the use of the BP in the CANDU reactor.