石墨慢化通道式熔盐堆有效缓发中子份额计算方法研究

有效缓发中子份额(βeff)是研究反应堆动力学特性的关键参数。在液态燃料熔盐堆(MSR)中,燃料流动引起缓发中子先驱核(DNP)在堆内的再分布,并使部分DNP在堆外回路衰变,从而导致βeff的计算方法与固态燃料反应堆不同。为评估石墨慢化通道式熔盐堆内燃料流动引起的反应性损失,研究缓发中子随燃料的流动行为,同时为堆设计和安全分析提供依据,分别基于解析方法和数值方法推导了计算βeff的数学模型,计算了熔盐实验堆(MSRE)在额定工况下的DNP损失份额和堆内DNP浓度分布,并分析了燃料在堆外流动时间和入口流量对βeff的影响。结果表明：两种方法均可对DNP行为提供合理描述;固定燃料在堆外流动时间,βeff随入口流量的增加而减小;固定入口流量,βeff随燃料在堆外流动时间的增加而减小,80 s后趋于稳定。     

Physical studies of minor actinide transmutation in the accelerator-driven subcritical system

The accelerator-driven subcritical system(ADS)with a hard neutron energy spectrum was used to study transmutation of minor actinides(MAs). The aim of the study was to improve the efficiency of MA transmutation while ensuring that variations in the effective multiplication factor(k_(eff)) remained within safe margins during reactor operation. All calculations were completed using code COUPLE3.0. The subcritical reactor was operated at a thermal power level of 800 MW, and a mixture of mononitrides of MAs and plutonium(Pu) was used as fuel.Zirconium nitride(ZrN) was used as an inert matrix in the fuel elements. The initial mass composition in terms of weight percentages in the heavy metal component(IHM)was 30.6% Pu/IHM and 69.4% MA/IHM. To verify the feasibility of this MA loading scheme, variations in k_(eff), the amplification factor of the core, maximum power density and the content of MAs and Pu were calculated over six refueling cycles. Each cycle was of 600 days duration, and therefore, there were 3600 effective full power days.Results demonstrated that the effective transmutation support ratio of MAs was approximately 28, and the ADS was able to efficiently transmute MAs. The changes in other physical parameters were also within their normal ranges.It is concluded that the proposed MA transmutation scheme for an ADS core is reasonable.     

ACP1000中MOX燃料组件管理初步研究

使用SCIENCE程序包对MOX燃料组件进行了初步设计和研究。在此基础上,对采用部分MOX燃料组件的ACP1000堆芯开展燃料管理研究,得到由全堆装载UO₂燃料组件向部分MOX燃料组件堆芯过渡的燃料管理方案,并对MOX燃料组件和部分MOX燃料组件堆芯的安全参数及其他重要参数进行分析和比较。分析结果表明,各种安全参数均满足设计要求,证明在ACP1000堆芯应用MOX燃料是可行的,并为进一步研究提供了参考。     

Assessment of fuel conversion from HEU to LEU in the Syrian MNSR reactor using the MCNP code

Assessment of fuel conversion from high enriched uranium (HEU) to low enriched uranium (LEU) fuel in the Syrian MNSR reactor was conducted in this paper. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed to assess the possibility of fuel conversion from 89.87% HEU fuel (UAl₄–Al) to 19.75% LEU fuel (UO₂). The first model showed that 347 fuel rods with HEU fuel were required to obtain a reactor core with 5.17 mk unadjusted excess reactivity. The second model showed that only 200 LEU fuel rods distributed in the reactor core like the David star figure were required to obtain a reactor core with 4.85 mk unadjusted excess reactivity. The control rod worth using the LEU fuel was enhanced. Finally, the third model showed that distribution of 200 LEU fuel rods isotropically in the 10 circles of the reactor core failed to convert the fuel since the calculated core unadjusted excess reactivity for this model was 10.45 mk. This value was far beyond the reactor operation limits and highly exceeded the current MNSR core unadjusted excess reactivity (5.17 mk).     

Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive reactivity effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors.     

钠冷氧化物燃料快堆嬗变MA研究

研究了次量锕系核素(MA)在钠冷氧化物燃料快堆中嬗变的基本物理特性。结果表明,MA核素加入堆芯燃料中后对堆芯动态参数和反应性反馈会产生显著的影响,如:βeff会有所减小、多普勒负反馈会显著减弱以及钠空泡反应性正反馈会显著增强。添加MA所带来的收益是燃耗反应性损失减小,且一定量的MA被嬗变掉,同时MA裂变也有相应的能量产出。MA嬗变的本质在于MA的焚毁,MA的焚毁比消耗与其所占全堆的裂变份额(包括由其转换的238Pu的裂变)成正比,为此相同MA裂变份额下的堆芯安全参数成为MA嬗变快堆设计的关键点。研究表明,堆芯小型化能够有效地减小堆芯的钠空泡反应性正反馈,同时对MA的焚毁比消耗影响较小。     

Safety analysis for core conversion (from HEU to LEU) of Pakistan research reactor-2 (PARR-2)

PARR-2 (Pakistan Research Reactor-2), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (High Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. The present report presents thermal hydraulic and safety analyses of PARR-2 using existing HEU fuel as well as proposed LEU fuel. Presently, the core is comprised of 90.2% enriched UAl₄-Al fuel. There are 344 fuel pins of 5.5 mm diameter. The core has a total of 994.8 g of U²³⁵. Standard computer code PARET/ANL (version 1992) (Obenchain, 1969) was employed to perform steady-state and transient analyses. Various parameters were computed, which included: coolant outlet, maximum clad surface & maximum fuel centerline temperatures; and peak power & corresponding peak core temperatures resulting from a transient initiated by 4 mK positive reactivity insertion. Results were compared with the reported data in Final Safety Analysis Report (FSAR) (Qazi et al., 1994). It was found that the PARET results were in reasonable agreement with the manufacturer's results. Calculations were also carried out for the proposed LEU core with two suggested fuel pin sizes (5.5 mm and 5.1 mm diameter with 12.6% & 12.3% enrichment, respectively). Comparison of the LEU results with the existing HEU fuel has been made and discussed.     

A minimum hot excess core design strategy for ESBWR

The Economic Simplified Boiling Water Reactor (ESBWR) is GEH’s next evolution of advanced BWR technology. There are 1132 fuel bundles in the core and the thermal power is 4500 MWt. As part of design simplification it uses natural circulation flow with no recirculation pumps or their associated piping. The control blades are the primary control mechanism to address the need for performing reactivity adjustments (using fine-motion drives) at or near rated steady state power. This introduces the potential for duty-related fuel failure, which has to be rigorously addressed as part of reliable design and operation. As means to mitigate this potential for duty-related fuel failure and also to support a simplified ESBWR operation, this study investigates the feasibility of a fuel cycle core design strategy. The objective is to design fuel bundles, and to use them for developing a core design, that minimizes (but does not eliminate) the use of control blades during operation. The reduction in use is envisioned in their number as well as movement in the core. In such a strategy, the effect of the burnable poison in the fuel (that largely drives the core reactivity) is enhanced, and operationally the control blades react modestly to maintain the core critical. While the logic is simple, challenges exist in developing such a design because it needs to balance the requirement for having enough blade inventory in the core to address design/operational constraints and uncertainties. The strategy is conceptualized as “minimum hot excess (reactivity)” design. It reduces the number of blades in the core during normal operation by 50% in comparison to a similar fuel cycle core design with regular inventory of control blades. Because of the increased burnable poison, the minimum hot excess core design strategy comes at a cost of fuel cycle efficiency. This cost is determined in terms of an increased enrichment for the fresh fuel batch fraction.     

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.     

Modeling reactivity devices for advanced CANDU reactors using the code DRAGON

Full core analysis of typical power reactors generally performed uses few group diffusion theory, it is necessary to generate beforehand, using a lattice code, the required few group cross-sections and diffusion coefficients associated with each region in the core.

For the ACR™ (Advanced CANDU Reactor), the problem is more complex because these reactors contain vertical reactivity devices that are located between two horizontal fuel bundles. The usual calculation scheme relies in this case on a 2D fuel cell calculation to generate the few group fuel properties and on a 3D supercell calculation for the analysis of the reactivity devices present in the core. Because of its complexity, the supercell calculations have usually been performed using simplified fuel geometries. The development of new geometry features in DRAGON and the availability of faster computers have made it possible to improve the 2D cell and 3D supercell models by using explicitly 3D assemblies of clusters to simulate the reactivity devices in CANDU reactors, including the ACR. These studies will thus improve the fine reactor core results by generating more accurate and appropriate reactor databases.

In this paper, we will review the lattice-cell/supercell calculation procedure using the code DRAGON by introducing a new supercell model. The use of such an explicit 3D geometry implies a very fine spatial mesh discretization that can generate a large number of regions leading to problems that cannot be solved by the collision probability (CP) method. The method of characteristics (MoC) is then the only alternative for such cases. A comparison of results using these two methods will also be presented for 3D models with a coarse mesh discretization.     

Nuclear fuel loading pattern optimisation using a neural network

An algorithm to optimise the fuel loading pattern (LP) in nuclear reactors was developed using an artificial neural network (ANN) to generate arrangements for the fuel in the core. The core parameters were calculated with the WIMS-D4 and CITATION-LDI2 codes, and the minimization of the maximum power peaking factor (FP_max) was used to choose the best arrangements. To verify the algorithm a PWR reactor with approximately 1/3 reprocessed fuel loaded was considered. The neutronic performance of the obtained arrangements and the efficiency of the implemented method were analysed. Several configurations were found for the core presenting better characteristics than the reference configuration adopted, so indicating the viability of the developed methodology. The algorithm was applied to a core considering part of the loading with reprocessed fuels, however this technique can be used for standard loadings.     

Detailed analyses of key phenomena in core disruptive accidents of sodium-cooled fast reactors by the COMPASS code

A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of key phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The key phenomena include (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, and (7) metal–fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several key phenomena are summarized. Simulations of GEYSER and THEFIS experiments were performed for dispersion and freezing behaviors of molten materials in narrow flow channels. In particular, the latter experiment using melt–solid mixture is also related to fundamental behavior of low energy disruptive core. CABRI-TPA2 experiment was simulated for boiling behavior of molten core pool. Expected mechanism of heat transfer between molten fuel and steel mixture was reproduced by the simulation. Analyses of structural dynamics using elastoplastic mechanics and fracture criteria were performed for SCARABEE BE+3 and CABRI E7 experiments. These two analyses are especially focused on thermal and mechanical failure of steel duct wall and fuel pin, respectively. The present results demonstrate COMPASS will be useful to understand and clarify the key phenomena of CDAs in SFRs in details.     

基于中子平衡研究增殖燃料实现CANDLE模式的最优配置

寿期内中子通量、核素浓度和功率分布的轴向形状均保持恒定(Constant Axial shape of Neutron flux,nuclide densities and power shape During Life of Energy produced,CANDLE)是实现原位增殖-焚烧(Breed-and-Burn,BB)模式的一种燃耗策略。CANDLE堆经易裂变燃料或外中子源进行点火,启动后由增殖燃料的燃烧实现自稳运行。若要CANDLE堆自稳运行于k_(eff)=1,必须对堆芯几何及燃料体积分数进行配置优化。最优配置方案可通过蒙特卡罗方法模拟CANDLE堆芯,根据有效增殖因子筛选得出。但该方法需耗费大量的计算时间,若采用1D模型近似模拟,并结合中子平衡方法进行分析,便可大幅节约计算时间,获得具有指导性意义的结果。本文将论证该方法的可行性,并应用该方法估算钠冷贫铀CANDLE堆半径在100 400 cm、燃料体积分数在35%60%变化时的最优配置。     

Transient and stability analysis of a BWR core with thorium–uranium fuel

The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket–seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal–hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO₂ core, even during transient conditions. The stability and transient analysis show that the thorium–uranium fuel can be operated safely in current BWRs.     

Reactor with a fast-resonance neutron spectrum, cooled by supercritical-pressure water with a bidirectional coolant flow scheme

A scheme for circulating coolant and cooling the core that has advantages over the designs of similar nuclear power systems is proposed for light-water reactors with supercritical coolant parameters and a fast-resonance neutron spectrum. A negative void coefficient of reactivity is obtained for the entire run of a fuel assembly without building a blanket. A more uniform distribution of the energy release over the core volume is achieved without using complicated fuel-enrichment schemes. The nonuniformity of the coolant temperature distribution at the core exit is decreased. The fuel assemblies operate with a much lower temperature drop over the core height. The core has a small reactivity excess on burnup and a BR of about 1, for which the most difficult operating regimes (flooding with cold water) can be handled with standard means (placement of absorbing organs of the safety and control system in ∼2/3 of the fuel assemblies). __________ Translated from Atomnaya énergiya, Vol. 100, No. 5, pp. 349–356, May, 2006.     

严重事故情况下环形燃料堆芯氢气产量分析

秦山Ⅱ期核电站反应堆堆芯采用环形燃料后,锆装量将增加约88%,在严重事故情况下,堆芯氢气产量的变化是一值得关注的问题。利用MELCOR程序模拟环形燃料堆芯,建立典型严重事故序列分析模型,分析结果表明:在堆芯熔化过程中,与传统棒状燃料堆芯相比,环形燃料堆芯氢气产量没有明显增加,使用环形燃料还能推迟事故进程,缓解事故后果。核电站采用环形燃料,不会增大氢气燃烧的风险。     

Effect of high-density fuel loading on criticality of low enriched uranium fueled material test research reactors

The effect of high-density fuel loading on the criticality of low enriched uranium fueled material test reactors was studied using the standard reactor physics simulation codes WIMS-D/4 and CITATION. Three strategies were considered to increase the fuel loading per plate: (1) by substituting the high-density fuel in place of low-density fuel keeping meat thickness and water channel width constant, (2) by substituting the high-density fuel in place of low-density fuel keeping fuel meat thickness fixed and optimizing the water channel width between the fuel plates and (3) by increasing the fuel meat thickness of fixed density fuel and optimizing the water channel width between the fuel plates. The fuel requirements for critical and first high power cores were determined in each case for higher fuel loadings per plate. It has been found that in the first case, core volume reduces with increasing fuel loadings per plate but requirement of fuel also increases. In the second and third case, core volume as well as fuel requirement decreases with increasing fuel loadings per plate. However in the second case, core volume reduces more rapidly than in case 3 with increasing fuel loadings per plate. Employing standard computer code PARET, steady state thermal hydraulic analysis of all these cores was performed. The thermal hydraulic analysis reveals that cores with higher densities and fixed water channel width are better from thermal hydraulic point of view and have fuel and clad temperatures within the acceptable limits. But the core with higher densities and optimum water channel width is a better choice in terms of core compaction, less ²³⁵U loading and higher neutron fluxes. Finally, the core was compacted in three steps to exploit the benefits of both types of cores. The strategy resulted in 36% reduction in the core volume, 50% increase in thermal neutron flux for irradiation and isotope production and a slight reduction in ²³⁵U loading. All this was achieved with acceptable peak clad and peak fuel centerline temperatures.     

A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO₂–ThO₂ fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO₂–ThO₂ fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO₂–UO₂ fuel pins are employed to achieve long-cycle length of ～4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods.     

Power up-grading study for the first Egyptian research reactor

In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4×4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5×5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2–10 MW) and different core coolant flow rates (450, 900, 1350 m³ h⁻¹) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5 — group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4×4-fuel basket and with the proposed 5×5-fuel basket up to 5 MW with the present coolant flow rate (900 m³ h⁻¹). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m³ h⁻¹ without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5×5) increases the excess reactivity of the reactor core than the present 4×4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.