Short-lived fission product release from UO2: Annular versus solid pellets

We have irradiated, in a “dual sweep” test assembly, one element containing solid UO₂ pellets with four, $\text{1 × 1}$ mm surface slots and one element containing annular UO₂ pellets with a central hole two mm in diameter. Data are reported for a linear power range of 17–57 kW/m; thermocouples monitored operating temperatures. He-2%H₂ carrier gas swept short-lived fission products from both elements past independent spectrometers for identification and measurement. The releases from the solid and annular fuel were power-dependent from 17–57 kW/m, with the annular fuel release up to eight times greater than that for solid fuel. The annular fuel also showed a larger released iodine inventory in fuel and sheath areas with access to the carrier gas.     

Thermal–hydraulic modeling of water/Al2O3 nanofluid as the coolant in annular fuels for a typical VVER-1000 core

In this paper, a thermal–hydraulic analysis of nanofluid as the coolant is performed in a typical VVER-1000 reactor with internally and externally cooled annular fuel. The fuel assembly for annular case with 8 × 8 arrays is considered for annular pin configuration. The considered nanofluid is a mixture composed of water and particles of Al₂O₃ with various volume percentages. The fuel rod is modeled using a CFD code. To validate the calculated results, the present results of solid fuel with nanofluid and pure water are compared with other studies which have been done with visual FORTRAN language, DRAGON/DONJON code, COBRA-EN code and the mentioned analytical approaches have been validated by comparing with the final safety analysis report (FSAR). The comparison of the calculated results shows that the results are in good agreement with other studies. Thus, the accuracy of the validation is satisfactory. Moreover, the temperature distributions of the fuel, clad and coolant are described for water/Al₂O₃ nanofluid in solid fuel and annular fuel. It is observed that as the concentration of Al₂O₃ nanoparticles increases, due to higher heat transfer coefficient of Al₂O₃ nanofluid, the temperature of the coolant is increased and the central fuel temperature is reduced. Thus, it improves margin from peak fuel temperature to melting. Finally, it is illustrated the use of the annular fuel instead of solid fuel in core of the reactor, security and efficiency of the nuclear power plant will be increased.     

Neutron absorber coatings for criticality control in annular tanks employed in spent fuel reprocessing

Computational analysis has been carried out to evaluate the effectiveness of neutron absorber coatings for criticality control in an annular tank used in fast reactor spent fuel reprocessing unit. The effect of composition, thickness and coating configuration for a given tank design and fuel solution concentration was evaluated on the basis of the multiplication factor (k_eff) calculated using the Monte Carlo N-Particle (MCNP) code. The neutron absorbers considered for the study were pure boron carbide (B₄C), B₄C/Ni–Cr combination and colmonoy. The effect of enriched boron was also analyzed. The results show that the coatings can enhance the storage capacity up to 30% for the annular tank studied.     

Effects of water based Al2O3, TiO2, and CuO nanofluids as the coolant on solid and annular fuels for a typical VVER-1000 core

In this paper, the effect of nanofluids as the coolant on solid and annular fuels for a typical VVER-1000 core is analysed. The considered nanofluids are various mixture composed of water and particles of Al₂O₃, TiO₂, and CuO. The fuel rod is modeled using a CFD code. To validate the calculated results, the present results of solid fuel with nanofluid and pure water are compared with other studies which have been done with visual FORTRAN language, DRAGON/DONJON code, COBRA-EN code and the mentioned analytical approaches have been validated by comparing with the final safety analysis report (FSAR). The comparison of the calculated results shows that the results are in good agreement with other studies. Thus, the accuracy of the validation is satisfactory. Radial and axial temperature distributions in various components of fuel are illustrated. Moreover, the temperature distributions of the fuel, clad and coolant are described for water based Al₂O₃, TiO₂, and CuOnanofluids in solid fuel and annular fuel. The results are compared with base fluid and it is concluded the nanoparticles of Al₂O₃have good properties in comparison with other nanoparticles. By using the nanofluids, the central fuel temperature is reduced and the temperature of the coolant is increased. In addition, by increasing the heated surfaces in annular fuel, the heat flux on these surfaces is reduced, the minimum departure from nucleate boiling ratio (MDNBR) margin is increased, and therefore the critical heat flux can be increased. Finally, it is concluded the use of the annular fuel instead of solid fuel and also the use of the nanofluids as coolant in the core of the reactor, security and efficiency of the nuclear power plant will be increased.     

Shut-down margin study for the next generation VVER-1000 reactor including 13 × 13 hexagonal annular assemblies

Shut-Down Margin (SDM) for the next generation annular fuel core of typical VVER-1000, 13 × 13 assemblies are calculated as the main aim of the present research. We have applied the MCNP-5 code for many cases with different values of core burn up at various core temperatures, and therefore their corresponding coolant densities and boric acid concentrations. There is a substantial drop in SDM in the case of annular fuel for the same power level. Specifically, SDM for our proposed VVER-1000 annular pins is calculated when the average fuel burn up values at the BOC, MOC, and EOC are 0.531, 11.5, and 43 MW-days/kg-U, respectively.     

双面慢化环形燃料反应堆物理实验

环形燃料零功率反应堆是首个双面慢化环形燃料作为核燃料的反应堆。本文采用周期法、落棒法获取环形燃料零功率反应堆的临界参数、控制棒价值、元件价值、含Gd元件的反应性效应等关键参数,对环形燃料零功率反应堆的物理性能进行实验研究,验证环形燃料反应堆堆芯物理设计计算程序。结果表明：根据外推过程确定堆芯临界装载环形燃料元件96根,实心燃料元件172根,此时k_eff为1.000 40,堆芯调节棒价值为-247.5 pcm,安全棒价值为-1 358.4 pcm;元件价值与理论值平均偏差为1.3 pcm,含Gd元件反应性效应与理论值平均相对偏差为8.8%。本文结果为环形燃料的工程化设计程序提供关键数据支撑。     

Two safety coefficients for a typical 13 × 13 annular fuel assembly

Recently, annular fuel rods are proposed for both PWRs and BWRs to achieve more average power density and therefore many neutronic as well as thermal-hydraulics calculations have been made to find more performance in the future reactors. Also, some safety margins are studied for the proposed uranium-nitride (and/or other fissile materials) annular pins. Our aim herein is to study two important safety coefficient of the annular fuel core. These are “prompt reactivity” and “power coefficients”, where all investigations are made using MCNP-5 code. Also, a thermal resistance model for annular fuel heat transferring were given and then we have calculated its (a case) thermal resistance numerically.     

Development of a thermal hydraulic analysis code for gas-cooled reactors with annular fuels

Gas-cooled reactors have been highlighted as a promising option for next generation reactor technology. A thermal hydraulic analysis code for gas-cooled reactors has been developed with a heat transfer model of a block element, which is solved implicitly with the helium energy equation. Validation was carried out through comparison with both experimental and analytical results. A computation module for annular fuel rods has been coupled to the code for comparative analyses of an annular fuel-based block element. At normal operation, the annular fuel shows 80 °C lower peak temperature than the solid fuel for the same power in Japan's high temperature engineering test reactor (HTTR), even though the pressure drop is higher in the annular fuel.     

LOCA工况下环形燃料元件内包壳外压屈曲试验研究

环形燃料一种安全高效的新型核燃料。为对环形燃料元件冷却剂丧失事故（LOCA）下整体受压失效形式的问题进行研究,将环形电加热棒、模拟芯块和试验件组装成试验装置,在空气环境中,以环形电加热棒外加热的方式,对环形燃料元件内包壳进行了外压屈曲试验,并将试验屈曲压力与Bresse?Bryan公式计算结果和特征值屈曲数值模拟分析结果进行了对比分析。结果表明：Bresse?Bryan公式计算结果除以安全系数m=2?5得到的结果高于试验结果而不够保守,试验结果分布于特征值屈曲数值模拟分析结果的1/5?1/3之间。本文结果可为环形燃料元件安全评价及后续工程化提供基础数据。     

百万千瓦级环形MOX燃料堆芯设计

对环形UO2燃料及环形MOX燃料组件参数的计算方法进行了研究。设计了包含193盒环形UO2和MOX燃料组件的混合型长周期(18个月)堆芯方案。对设计的堆芯的重要物理参数进行了分析,并对各循环进行了燃耗计算。结果表明,装载约30%MOX组件的堆芯可在百万千瓦功率下实现长周期换料。堆芯从初装载可安全过渡到平衡循环,各循环的重要物理参数均满足设计要求,说明设计的堆芯及燃料管理方案是安全可行的。     

环形燃料棒束再淹没行为实验研究

环形燃料是一种可在维持或提升安全裕度的前提下大幅提高反应堆经济效益的新型压水堆燃料,由于其双面冷却的特点,环形燃料在LOCA再淹没阶段的热工水力行为与传统实心燃料存在显著差异。现有关于环形燃料再淹没行为的实验研究鲜有报道。本研究基于自主设计的高温环形电加热棒建立了环形棒束再淹没实验装置,开展了3×3环形棒束底部再淹没实验研究,探究了环形棒束再淹没典型物理过程及不同工况下再淹没关键参数的变化规律。结果表明,环形棒束再淹没物理过程与传统实心棒束类似,且内外通道的骤冷前沿推进和传热模式变化趋于同步。在同一时刻下,环形棒内外壁面间存在温度梯度。骤冷前沿推进速度随再淹没速度和过冷度的增大而增大,随峰值包壳温度和线功率密度的增大而减小。此外,定位格架在低流速、低过冷度与高壁温工况下能显著提升下游的骤冷前沿推进速度。     

Fission product precipitates in irradiated uranium carbonitride fuel

Austenitic steel-cladded uranium carbonitride fuel pins were irradiated in the BR2 up to 6.4% burnup. A cross-section of the pin RV 24 with the fuel composition UC_0.86N_0.09O_0.05 was prepared for X-ray microanalysis of the fission product precipitates. Rare-earth oxide and U(Mo,Tc)C₂ phases were observed in the whole fuel region. Bright phases present in annular rings of the outer fuel zone were identified as U₂(Tc, Ru, Rh)C₂. Alkaline-earth oxide and U–Pd–Ni phases were shown in the fuel-cladding gap. The rare-earth and alkaline-earth fission products extracted the oxygen from the fuel matrix which became nearly oxygen free. The formation of nitrides could not be detected.     

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.     

秦山二期核电站环形燃料堆芯卡轴及弹棒事故分析

环形燃料是一种采用双层包壳和环形芯块内外冷却的新型压水堆燃料,与传统棒状燃料相比,双包壳结构有效增加了燃料传热面积和减薄了芯块厚度,使其在事故工况下具有更好的安全性能。以秦山二期核电站作为参考电站,建立了装载环形燃料的核电站计算模型,研究在卡轴事故和弹棒事故下采用环形燃料的核电站的响应,并与相应工况下棒状燃料堆芯的计算结果比较。结果表明,与棒状燃料相比,核电站在采用环形燃料后安全裕度有明显的提高。     

~(137)Cs源燃料棒芯块间隙检测装置设计

设计了一种使用~(137)Csγ放射源检测新型燃料棒内芯块间隙的装置。该装置除具备241Am源检测装置功能外,还可清晰辨别实心芯块和环形芯块区域,以及辨别燃料棒内的芯块垫、弹簧、支撑管等。使用1.5Ci的~(137)Cs放射源,装置检测速度6 m/min、置信度95%时,设备的燃料棒长度检测精度为4mm/1m,间隙检测精度为±0.13 mm。装置运行稳定性良好,精度满足检测需求。     

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。     

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.     

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li₂₀Sn₈₀ coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.     

Development of a ϕ20 mm annular centrifugal contactor for the hot test of the total TRPO process

Centrifugal contactors have many advantages and are favored for reprocessing of spent nuclear fuel and partitioning of high-level waste. A new ?20 mm annular centrifugal contactor for laboratory scale has been developed for the hot test of the total TRPO process. A modular design, a multi-stage group housing design and ceramic bearings are adopted in the new centrifugal contactor. A sampling system and a rotor speed acquisition and monitoring system for ?20 mm annular centrifugal contactors have been developed for tests in the hot cell. The hydraulic performance of both a single-stage centrifugal contactor and a three-stage cascade has been studied. The nominal throughput of the single-stage centrifugal contactor can reach 13.5 l/h under suitable operation conditions for H₂O–kerosene system. The hydraulic performance of the three-stage cascade is also good.     

环形燃料热工水力性能分析程序开发及验证

本工作开发了环形燃料子通道分析程序SAAF。采用SAAF计算了西屋公司四环路压水堆所用环形燃料组件的热工水力性能,并与VIPRE-01的计算结果进行比较。结果表明,SAAF与VIPRE-01的计算结果符合较好,SAAF可用于环形燃料热工水力设计分析。