Pressure Drop Performance of Fuel Pin Bundle with Spiral Wire Spacer

The pressure drop characteristics of a JEFR type fuel pin bundle were obtained from hydraulic tests. The coefficient of drag attributable to the spiral wire spacer wound round each fuel pin, as defined by de Stordeur, was found to be approximately 0.30 for the hexagonal lattice arrangement adopted. The coefficient is independent of spiral wire pitch, which ranged from 90 to 260 mm.

The pressure drop to be expected in a fuel pin bundle with spiral wire spacer, such as used in current fast breeder applications, can be satisfactorily estimated by using the coefficients reported.     

Development of a Method for Evaluation of Pin-wise Power Distribution in Fuel Assemblies of Fast Reactors

In the design of fast reactor core with higher burnup and higher linear power, prediction accuracy of burnup history of fuel pin should be upgraded so as to assure fuel integrity without extra design margin under increased neutron fluence and burnup. A method is studied to predict fuel pin-wise power and its burnup history in fast reactors accurately based on an analytic solution of diffusion theory equation on hexagonal geometry with boundary condition from core calculation by finite-differenced diffusion calculation code. The present method is applied to a fast reactor core model, and its accuracy in predicting fuel pin power is tested. The result is compared with the reference solution by the finite difference calculation with very fine mesh. It is found that the present method predicts the power peaking factors in fuel assemblies accurately. The fuel pin-wise nuclide depletion calculation is also done using neutron fluxes for each fuel pin. The result shows that the fuel pin-wise depletion calculation is very important in predicting the burnup history of the fuel assembly in detail.     

超临界水冷堆环形燃料组件核热耦合分析

在超临界水冷堆预概念设计中,组件设计是十分重要的,将影响堆芯性能。超临界水冷堆中水密度变化剧烈的特性要求必须进行核热耦合分析。从中子学及热工性能角度,使用三维核热耦合程序对环形燃料组件进行了优化设计。应用中子学计算程序FENNEL-N对环形燃料组件进行三维扩散计算,可得到组件内单棒功率分布,应用热工计算程序SUBSC对组件进行子通道分析。在计算过程中,分析了燃料棒间距及燃料棒与组件壁盒之间的间隙对组件性能的影响。计算结果显示,增大棒间距和棒壁间隙能提高组件k_inf,但会增大组件内功率峰因子;子通道受热不均匀性对组件热工性能影响较大,通过加入定位格架的方式能展平冷却剂出口温度,降低最大包壳温度。对环形燃料组件的安全分析表明,从中子学角度该组件是安全的。     

Reactivity Worth of Fuel Elements

Reactivity worths of fuel elements were measured in the Ozenji Critical Facility (OCF) and analyzed with three group perturbation method. The result shows that the worth of one single fuel pin can be well predicted by calculation over a very wide range of the core spectrum, namely from a lattice of 2.5% enrichment and 0.43 volume ratio to that of 1.5% enrichment and 3.5 volume ratio.

The analysis indicates the importance of thermal neutron flux peaking remaining after the removal of a fuel pin. Only by incorporating this effect can the reactivity worth of a fuel pin be correctly evaluated. In the present study, the neutron spectrum in the water hole where the peaking occurred was assumed to be the same as in the reflector. The reflector spectrum seems to provide better agreement with experiment than the core spectrum. Validity of the analysis was extended to a bundle of sixteen fuel pins by measuring the reactivity worths of bundles of fuel pins as well as the thermal neutron flux distributions. One dimensional diffusion calculations were employed throughout the analysis.     

快堆燃料组件内精细功率分布的计算

堆芯热通道因子是堆芯热工设计及安全分析的一项重要参数,确定热通道因子需用中子学计算给出较准确的燃料组件内元件棒功率分布。在三维六角形几何节块扩散理论基础上,使用多项式重构的方法计算节块内中子通量密度分布和功率密度分布。针对快堆六角形燃料组件的特点,用小六角形积分的方法计算组件内元件棒功率,得到组件内各元件棒功率分布。在NAS程序基础上,编制了元件棒功率分布计算模块NAS PIN。通过与蒙特卡罗程序的校验可发现,二者计算结果符合较好,计算精度可满足工程设计的需要。     

Irradiation Performance of Fast Reactor MOX Fuel Assemblies Irradiated to High Burnups

Mixed oxide fuel assemblies (MFA-1 and MFA-2 assemblies) were irradiated in the fast flux test facility to evaluate the irradiation performance of fast reactor core fuels at high burnups and high fast neutron fluences. The MFA-1 and MFA-2 assemblies achieved respective peak pellet burnups of 147 and 162GWd/t, and resisted to respective peak fast neutron fluences (E > 0:1 MeV) of 21:4 _ 1026 and 23:8 _ 1026 n/m2, without any indication of fuel pin breaching. Structural components of these assemblies were made of modified type 316 stainless steel and 15Cr-20Ni base advanced austenitic stainless steel. Postirradiation examinations of these assemblies revealed dimensional changes of fuel pins and assembly ducts due to irradiation-induced void swelling and irradiation creep, and fuel cladding local oval distortions due to bundle-duct interaction (BDI). The swelling resistance of 15Cr-20Ni base advanced austenitic stainless steel fuel pin cladding was almost the same as that of the modified type 316 stainless steel cladding, while the assembly duct of the former material had a slightly higher swelling resistance than that of the latter material. Analyses of fuel pin bundle deformations indicated that these assemblies likely mitigate BDI mainly by fuel pin bowings and cladding oval distortions.     

田湾核电站TVS-2M先导组件堆芯热工水力分析

田湾核电站一号机组于第5燃料循环装入6组TVS-2M先导燃料组件,并将经历从第5燃料循环到第8燃料循环4年的堆内运行。本文通过对先导燃料组件堆芯热工水力分析,堆芯运行实际试验测量以及组件变形检查,验证了热工水力设计程序计算模型的合理性以及计算结果与试验结果的符合性。结果表明,TVS-2M燃料组件与AFA燃料组件具有良好的相容性,从而证实了过渡循环条件下反应堆运行的安全性和可靠性。     

Sensitivity analysis of fuel pin failure performance under slow-ramp type transient overpower condition by using a fuel performance analysis code FEMAXI-FBR

The FEMAXI-FBR is a fuel performance analysis code and has been developed as one module of core disruptive evaluation system, the ASTERIA-FBR. The FEMAXI-FBR has reproduced the failure pin behavior during slow transient overpower. The axial location of pin failure affects the power and reactivity behavior during core disruptive accident, and failure model of which pin failure occurs at upper part of pin is used by reflecting the results of the CABRI-2 test. By using the FEMAXI-FBR, sensitivity analysis of uncertainty of design parameters such as irradiation conditions and fuel fabrication tolerances was performed to clarify the effect on axial location of pin failure during slow transient overpower. The sensitivity analysis showed that the uncertainty of design parameters does not affect the failure location. It suggests that the failure model with which locations of failure occur at upper part of pin can be adopted for core disruptive calculation by taking into consideration of design uncertainties.     

Cell geometry effect on equilibrium state of nuclear reactors

The effects of the cell configuration on core performance for a liquid sodium cooled MOX fuel type fast reactor and a PWR type thermal reactor are investigated. In this study our equilibrium cell iterative calculation system (ECICS) are used in order to obtain consistent neutron spectra, one-group constants and nuclide number densities at the nuclear equilibrium state. The fuel pellet diameter and the pin pitch are changed to evaluate their core characteristics. The distinction of reactors at the equilibrium state appears clearly by means of the change in the cell geometrical design.     

Fuel transient deformation

Cladding strains resulting from fuel-cladding mechanical interaction in a transient tested fresh fuel pin are assessed against laboratory measurements of high-temperature creep and hot pressing of mixed-oxide fuel pellets.A fuel pin containing nine different fuel sections with varying fuel pellet geometry and density was transiently tested in a MARK IIIA flowing sodium loop under conditions typical of a 1$/s overpower transient. Post-test cladding strain measurements indicated that the largest strains were generated by solid pellets with small gaps while large gap annular pellets generated the smallest strains.High-temperature creep and hot pressing tests on mixed-oxide fuel have been performed at temperatures up to 2600°C. The results indicate that at temperatures above 2300°C, an additional component of creep is operative; while the densification due to hot pressing was considerably less than expected by extrapolating the low-temperature behavior.Both the in- and out-of-reactor data suggest that fuel creep into the center void or hole of a fuel pin is a more effective means of reducing fuel-cladding stress than densification by hot pressing into fuel pellet porosity.     

径向倒料式驻波堆堆芯概念设计

本工作在综合分析日本CANDLE堆和美国TerraPower公司设计的TP-1堆的基础上,提出沿径向倒料的驻波堆堆芯设计初步方案,通过高、低富集度组件在堆芯的混合布置展平功率,降低了堆芯的功率峰因子和组件的最大燃耗深度。通过倒换料,实现了增殖 燃耗波的传递。为能有效地利用行波堆增殖产生的易裂变核素,采用更换包壳的新技术,实现核燃料的持续利用。     

Analysis of Cell Configuration Effects on Nuclear Characteristics in a Lead-Bismuth Cooled Core for TRU Incineration

Sensitivity of the core characteristics to the fuel pin cell parameters change is analyzed for a lead-bismuth cooled reactor to incinerate transuranic nuclides. The pitch-to-diameter ratio is changed for a parametric study to investigate the effects of the coolant-to-fuel ratio. Not only the Zr-based fuel of TRU+Zr but also the Th-based fuel of TRU+Th+Zr is considered in order to investigate the sensitivity of nuclear characteristics of the fuel pin cell to neutron energy spectrum as well as effects of the fuel type on the core performance. For the sensitivity analyses, the neutron spectrum, the criticality performance parameters, and the non-fissile actinides destruction factor are evaluated. The obtained results clarify the unique property of nuclear characteristics of the fuel pin cell and give some useful information for design optimization of a lead-bismuth cooled reactor for TRU transmutation.     

等效天然铀入堆辐照试验中子学性能分析

本文基于重水堆堆芯核设计程序系统,计算分析了装入等效天然铀（NUE）燃料的试验堆芯的中子学性能。对选择两个燃料通道进行入堆试验的方案进行了论证分析,通过与相应的设计限值以及全天然铀（NU）燃料堆芯中子学性能的比较,检验了实际入堆NUE燃料的中子学等效性。研究表明,实际入堆NUE燃料满足燃料的等效性要求,两个NUE燃料通道入堆试验方案从核设计角度是可行的,堆芯安全性不受影响。     

Fuel Pin Behavior under Slow-Ramp-type Transient-Overpower Conditions in the CABRI-FAST Experiments

In the CABRI-FAST experimental program, four in-pile tests were performed with slow-power-ramptype transient-overpower conditions (called hereafter as “slow TOP”) to study transient fuel pin behavior under inadvertent control-rod-withdrawal-type events in liquid-metal-cooled fast breeder reactors. The slow TOP test with a preirradiated solid-pellet fuel pin under a power ramp rate of approximately 3%Po/s was realized as a comparatory test against an existing test in the CABRI-2 program where approximately 1%Po/s was adopted with the same type of fuel pin. In spite of the different power ramp rates, the evaluated fuel thermal conditions at the observed failure time are quite similar. Three slow TOP tests with the preirradiated annular fuel resulted in no pin failure showing a high failure threshold. Based on posttest examination data and a theoretical evaluation, it was concluded that intrapin free spaces, such as central hole, macroscopic cracks, and fuel-cladding gap, effectively mitigated the fuel cladding mechanical interaction. It was also clarified that cavity pressurization became effective only in the case of a very large amount of fuel melting. These CABRI-FAST slow TOP tests, in combination with the existing CABRI and TREAT tests, provided an extended slow TOP test database under various fuel and transient conditions.     

Design research of small long life CANDLE fast reactor

The feasibility of a small long life fast reactor with CANDLE burn-up concept was investigated. It was found that a core with 1.0 m radius and 2.0 m length can bring about CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead–Bismuth is used as coolant. From equilibrium analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year, which easily permits a long core life design. The averaged core discharged fuel burn-up is about 40%. For better understanding of the effect of the coolant to fuel volume ratio, comparison was made among five cases. In these cases the coolant channel radii were different from one case to another, while fuel pin pitch was fixed. Comparisons were also made with a fixed coolant channel radius and different fuel pin pitches. A simulation of core operation is implemented and the results show that the present design can establish the long time steady CANDLE burn-up successfully without a burn-up control mechanism.     

遗传算法在堆芯燃料管理中的应用

应用遗传算法编制了核反应堆堆芯燃料管理优化计算程序。在遗传算法程序中，选用了性能良好的编码方法，优化了遗传算子各个参数，使遗传算法程序的计算质量和效率有明显的提高。从计算结果看，遗传算法能够获得良子的堆芯燃料布置。     

AP1000乏燃料池非能动冷却系统事故后冷却能力分析

AP1000乏燃料池冷却系统采用了先进的非能动设计理念,事故后以池水升温-沸腾的方式带走衰变热,并通过持续的非能动安全补水保证乏燃料安全。对AP1000乏燃料池冷却系统的事故后冷却能力进行分析发现,在核电厂正常换料工况和应急整堆芯卸载工况下,安全水源重力注水能保证事故后72 h内乏燃料安全;在核电厂正常整堆芯换料过程中应等待约56 h,以保证非能动安全壳冷却水箱可为乏燃料池补水,确保堆芯和乏燃料池安全。长期补水可以通过预留的安全接口持续进行。补水手段事故后有效,仅需操纵员有限干预。相对传统乏燃料池冷却系统设计,AP1000能更好地应对冷却丧失的事件。     

中国先进研究堆标准燃料组件堆外水力稳定性试验

中国先进研究堆（CARR）标准燃料组件由滚压在两块侧板上的21块燃料板组成。堆外水力试验的目的是考验在水力冲刷条件下燃料组件的结构稳定性。试验件是按照正式产品制造工艺制造的贫铀组件,试验平均流速为12m/s,是满功率运行流速的120％。先后试验了2个组件,第1个组件试验60d,是满功率运行时间的120％,试验后观察到固定下定位梳的销钉松动,下定位梳严重磨损了燃料板;工艺改进后制造的第2个组件试验120d,是满功率运行时间的240％,试验表明,第2个组件结构完整。试验中对组件结构稳定性和燃料板腐蚀性能,诸如组件的压差、燃料板振动、包壳表面腐蚀深度等进行了研究。     

Statistical evaluation of pin gaps in fuel clusters for fugen (II)

Fuel pin gaps of Fugen fuel assemblies deviate statistically from their nominal value due to manufacturing and assembling tolerances which influence the thermal and hydraulic characteristics of the reactor core. For assurance of the minimum fuel pin gap, an analytical method of evaluating the reliability of spacer gauge tests applied to selected fuel pin gaps arrayed within a Fugen fuel assembly is discussed where a computer program STGAP is utilized.Correlations among the thickness of a spacer gauge, the reliability of the test and the rate of rejecting fuel assemblies whose pin gaps all satisfy the minimum design criterion are discussed in connection with the optimum gauge thickness for a given realiability level of the test. Sample calculation shows that fuel subassemblies installed in a Fugen reactor core have the overall reliability level of 99.9954% at the beginning of fuel life.     

小型自然循环钠冷堆堆芯初步设计研究

提出了一种适用于分布式发电系统的小型自然循环钠冷堆AMTEC系统。通过对堆芯的临界计算和热工水力分析,研究了堆芯燃料装载量不变情况下,芯块半径、燃料棒长度和圈数对堆芯有效增殖因数k_eff、堆芯压降和传热的影响。同时分析了不同额外停堆裕量下,B₄C吸收层厚度和堆芯初始剩余反应性随燃料棒圈数的变化关系。计算结果表明：保持堆芯当量直径和冷却剂通道总截面积不变的情况下,减少燃料棒圈数和活性区长度不仅可增加k_eff,且能降低堆芯压降;为提高额外停堆裕量需增加吸收层厚度,但降低了堆芯初始剩余反应性,不利于电厂的经济性。