包覆燃料颗粒应力的有限元分析

高温气冷堆的燃料元件由包覆燃料颗粒弥散在石墨基体中组成。在反应堆运行过程中,辐照及各复杂的物理化学反应产生的应力会使包覆燃料颗粒发生破损,对包覆燃料颗粒进行应力分析是评价燃料元件和反应堆运行安全性能的主要内容之一。本文基于压力壳模式,主要考虑内压作用下的球形壳层应力及包覆燃料颗粒的非球形因素,用有限元法对应力进行了分析。     

碳化硅基新型包覆燃料颗粒的设计及制备

TRISO型包覆燃料颗粒可将核裂变产生的气体、固体裂变产物束缚在燃料颗粒内部,是高温气冷堆安全性的重要保障。为满足未来超高温气冷堆在更高温度及更高燃耗条件下对燃料元件的要求,需对传统TRISO颗粒进行优化和改进。基于包覆颗粒的破损机制,设计了两种SiC基新型包覆颗粒,一种采用疏松SiC层替代疏松热解炭层,包覆层由内而外依次为疏松SiC层、内致密热解炭层、致密SiC层、外致密热解炭层;另一种为全SiC包覆结构,包覆层由内而外依次为内层疏松SiC层、SiC过渡层、外层致密SiC层。根据结构设计,采用流化床化学气相沉积法实验探索了疏松SiC的形成机制及包覆工艺条件,并利用SEM、XRD等进行材料分析,最终成功实现了两种新型包覆颗粒的大规模制备。更进一步,提出了全SiC基燃料元件的概念,并制备了球形和柱形全SiC基模拟燃料元件。     

Measurements of Doppler Effect of Coated Particle Fuel Rod in SHE-14 Core Using Sample Heating Device

Reactivity decrease due to temperature rise of a single fuel rod sample was measured in the SHE-14 core using a sample heating device with purpose to verify the calculation accuracy of the Doppler effect for resonance neutron absorption in a Very High Temperature Reactor. The fuel rod sample was a stuck of coated particle fuel compacts containing 4% enriched UO₂ kernels, and it was heated up to about 700°C in a sample heating tube which was placed along the axis of the core.

The difference of reactivity decrease between the two same size samples of fuel rod and graphite rod due to temperature rise can be interpreted as the increased resonance neutron capture of ²³⁸U, i.e. Doppler effect. The SRAC code system was applied to the Doppler effect calculation where the collision probability method was used in the cell calculation and the one-dimensional, multi-group diffusion approximations were adopted in the core calculation. The results gave a ratio of the calculated to the measured Doppler effect of 0.93.     

包覆燃料颗粒制备的自动化控制系统设计与研制

针对高温气冷堆包覆燃料颗粒生产逐渐规模化大批量发展的趋势,原制备工艺的手动控制体系已不能适应,需发展现代化工业级别的包覆燃料颗粒制备自动化控制系统。针对TRISO型包覆燃料颗粒4层连续包覆工艺进行分析,将包覆炉系统划分为5个子系统,将整个包覆过程分解为9个操作状态,提出建立分布式控制系统（DCS）自动化控制系统的思路。根据对包覆颗粒制备工艺的严格要求提出了控制系统设计原则,包括连锁控制、安全可靠、集成规范、实用易用、开放和易更新原则,并在具体建设过程中实现了这些原则要求,建立起一套完整的包覆燃料颗粒制备工艺自动化控制系统。该系统在我国高温气冷堆示范工程项目辐照样品的生产中投入运行,经实践检验,证明该系统可较好地实现包覆燃料颗粒制备工艺控制,满足工厂规模的生产要求。     

包覆燃料颗粒生产线压力平衡分析及系统优化

高温气冷堆使用的TRISO型包覆燃料颗粒采用流化床化学气相沉积法制备。实验研究发现,在包覆燃料颗粒的生产过程中,化学气相沉积过程的压力显著影响包覆燃料颗粒的性能指标。研究了包覆过程中反应压力的变化对包覆层的显微形貌及密度的影响规律。基于对化学气相沉积反应压力的控制需求,进行生产线全系统压力监测及分析,获得整个生产线压力平衡及稳定操作的值。对生产线进行了系统优化,提出从后端调节压力的方法,通过系统的压力平衡,实现了对反应压力的有效控制,保证了生产过程稳定性。研究结果指出,全系统压力平衡分析是一种有效的生产线评估手段,可用于故障分析排除、分级连锁设计以及系统优化等。     

Research and Development of HTTR Coated Particle Fuel

The coated particle fuel has been developed within a framework of the HTTR (High Temperature engineering Test Reactor) Development Program at the Japan Atomic Energy Research Institute. The HTTR fuel is a prismatic block type containing TRISO-coated U0₂ particles. Research and development on the fuel has been progressed in three categories; a work for fuel production technology, a proof test of fuel performance and a safety-related research. In the present report the concept and outline of the fuel in the HTTR design are firstly described, and then fuel fabrication technology including recently developed methods for improving fuel quality is followed. Tests for proving fuel performance have been carried out extensively on the reference fuel of the HTTR design by irradiation in an in-pile gas loop and capsules, and typical results are presented in this report. Concerning the safety-related research, fuel failure and ¹³⁷Cs release at abnormally high temperature are described.     

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.     

Effects of Fuel Oxidation and Dissolution on Volatile Fission Product Release under Severe Accident Conditions

The release of volatile fission products from high-burnup UO₂ fuel was examined in a steam atmosphere under severe accident conditions as a part of the VEGA program. The effects of fuel oxidation and dissolution were totally evaluated, by comparing the results with those from previous inert, hydrogen and steam atmosphere tests. It was shown that the oxidation of UO₂ to UO_2+x by steam generally enhances Cs and Kr release. However, the enhancement becomes smaller above the melting temperature of Zircaloy, about 2030 K, likely due to reduction of UO_2+x by molten Zircaloy. The burst release of Cs occurs above about 2300K in the hydrogen atmosphere, while the release rate does not increase so significantly for the examined temperature range (<2800 K) in the steam atmosphere. Analysis of the hydrogen atmosphere test showed that fuel dissolution is apparently connected with the burst release and that a large fraction of Cs is quickly released from the dissolved fuel above 2300 K. It is considered that the fuel dissolution rate in the steam atmosphere is about 1/1000 of that in the hydrogen atmosphere.     

Analysis of fuel temperature effects on reactivity of light water reactor fuel assemblies by using MVP-2 adopting an exact resonance elastic scattering model

Using the continuous-energy Monte Carlo code MVP-2 adopting a resonance elastic scattering model considering the thermal motion of a target nucleus (the exact model) for major heavy nuclides, analysis of fuel temperature effects on reactivity of mockup UO₂ and MOX fuel assemblies for light water reactors was performed, and the results were compared with those of the conventional asymptotic model. A base condition was a hot operating condition with an in-channel void fraction of 40% and fuel temperature of 520 ℃ for the BWR fuel assemblies and a hot zero-power condition with fuel temperature of 284 ℃ for the PWR fuel assemblies. The fuel temperature of a high-temperature condition was 1500 ℃ for both types of assemblies. The calculated results showed that the exact model made the neutron multiplication factors at the high-temperature condition lower by ?220 to ?440 pcm (10^?5 Δk) and the Doppler reactivity between the base- and high-temperature conditions more negative by 7% to 10% compared with those obtained by the asymptotic model. The energy-dependent reaction rates of capture and ν-fission were also analyzed to study the detail mechanism in the effect of the exact model on the assembly reactivity.     

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.     

Investigation of thermal neutron scattering data for BeF2 and LiF crystals

The influences of thermal neutron scattering data for BeF₂ and LiF crystals on molten salt reactor physics are investigated in this work. Based on the structure parameters of BeF₂ and LiF, the coherent scattering for both crystals is added to NJOY source code. The ENDF6 format thermal neutron scattering sub-libraries for both crystals are evaluated with their phonon spectra using LEAPR; the ACE format data are produced by NJOY subsequently. Finally, the effect of thermal neutron scattering of BeF₂ and LiF crystals on k _eff and spectrum are investigated. The result shows that thermal neutron scattering for bound state of BeF₂ and LiF influence k _eff and spectrum obviously. The elastic scattering cross section for bound state of crystals is smaller than free atom; it makes k _eff decrease (1%–2%) and spectra be hardened. The higher temperature the bound state has, the smaller coherent elastic scattering cross section it gets; therefore, k _eff decreases with temperature. It is suggested that the thermal neutron data of LiF and BeF₂ should be taken into account for molten salt reactor.     

Recovery of neutron-irradiation-induced defects of Al2O3, Y2O3, and yttrium-aluminum garnet

SiC fiber-reinforced SiC matrix composites (SiC_f/SiC) are considered as one of the candidates for blanket materials in future fusion reactors and as an advanced fuel cladding material for next-generation fission reactors. Generally, the densification of SiC needs sintering additives and oxides such as Al₂O₃, Y₂O₃, and yttrium-aluminum garnet (YAG, Y₃Al₅O₁₂), which are frequently added to SiC. However, the effects of neutron irradiation on sintering additives are still unclear. In this study, we performed the neutron irradiation of Al₂O₃, Y₂O₃, and YAG at fluences up to 2.0–2.5 × 10²⁴ n/m² (E > 0.1 MeV) at 60–90 °C. The isochronal recovery of the macroscopic volume of Al₂O₃ against annealing temperature showed smooth and continuous shrinkage at a temperature of up to 1200 °C, and the volume slightly increased above that temperature. In contrast, the volume of Y₂O₃ showed quick shrinkage at the low temperature range, and slower and smooth recovery was observed up to ～1100 °C. In the case of YAG, the recovery of volume occurred in a step-wise manner at 600–750 °C, and continuous shrinkage occurred at temperatures lower and higher than that temperature range. The activation energies for the macroscopic volume recoveries of three oxides were obtained from the Arrhenius plots of the rate coefficients. Two-stage recovery was observed for Al₂O₃, whereas more complicated recovery processes were suggested for Y₂O₃ and YAG.     

Determination of 79Se and 135Cs in Spent Nuclear Fuel for Inventory Estimation of High-Level Radioactive Wastes

⁷⁹Se and ¹³⁵Cs are long-lived fission products and are found in high-level radioactive waste (HLW). The estimation of their inventories in HLW is essential for the safety assessment of geological disposal, owing to their mobility in the strata. In this study, the amounts of ⁷⁹Se and ¹³⁵Cs in a spent nuclear fuel solution were measured. About 5 g of irradiated UO₂ fuel discharged from a commercial Japanese pressurized water reactor (PWR) with a burn-up of 44.9 GWd/t was sampled and dissolved with 50mL of 4M nitric acid in a hot cell for 2 h. After Se and Cs were chemically separated, the amounts of ⁷⁹Se and ¹³⁵Cs in the spent nuclear fuel solution were measured by inductively coupled plasma quadrupole mass spectrometry (ICP-QMS). The amounts of ⁷⁹Se and 135Cs were 5:2 ± 1:5 and 447 ± 40 g/MTU, respectively. The results presented in this study, which are the first postirradiation experimental data in Japan, showed good agreement with those obtained by the ORIGEN2 code using the data library of JENDL-3.3.     

Transport Calculations of MOX and UO2 Pin Cells by the Method of Characteristics

A new transport theory code for two-dimensional calculations of both square and hexagonal fuel lattices by the method of characteristics has been developed. The ray tracing procedure is based on the macroband method, which permits more accurate spatial integration in comparison to the equidistant method of tracing. The neutron source within each region is approximated by a linear function and linearly anisotropic scattering can be optionally accounted for. Efficient new techniques for both azimuthal and polar integration are presented. The spatial discretization problem in case of P ₁-scattering has been studied. Detailed analyses show that the P ¹-scattering in case of regular infinite array of fuel cells is significant, especially for MOX fuel, while the transport correction is inadequate in case of real geometry multi-group calculations. Finally, the complicated nature of the angular flux in MOX and UO₂ fuel cells is demonstrated.     

Concentrations of CsCl and SrCl2 from a Simulated LiCl Salt Waste Generated by Pyroprocessing by Using Czochralski Method

Separation of CsCl and SrCl₂ from LiCl was carried out by using a separation technology, the Czochralski crystallization method. It was experimentally confirmed that Cs as well as Sr could be separated simultaneously from a LiCl molten salt by the suggested crystallization process without any additive or adsorption medium. The concentrations of Cs and Sr in LiCl decreased from 1.81 and 4.18 wt% in the initial salt to minimum values of 114 and 36 ppm in the grown LiCl crystal, respectively. The separation mechanism of Cs and Sr is described by the solubility difference of the solutes between the molten and solid states. It is expected that the total amount of salt waste will decrease drastically, because most of LiCl could be recovered by recycling with an electroreduction process.     

Neutron cross section sensitivity analysis of UO2 and MOX core physics experiments on light water reactors

The perturbation theory based on the transport calculation has been applied to study sensitivity of neutron multiplication factors (k_eff's) to neutron cross sections used for the reactivity analysis of UO₂ and MOX core physics experiments on light water reactors. The studied cross sections were neutron capture, fission and elastic scattering cross sections, and a number of fission neutrons, ν. The obtained sensitivities were multiplied to relative differences in the cross sections between JENDL-4.0 and JENDL-3.3 in order to estimate the reactivity effects. The results show that the increase in k_eff, 0.3%Δk/kk′, from JENDL-3.3 to JENDL-4.0 for the UO₂ core is mainly attributed to the decreases in the capture cross sections of ²³⁸U. On the other hand, there are various contributions from the differences in the cross sections of U, Pu, and Am isotopes for the MOX cores. The major contributions to increase in k_eff are decreases in the capture cross sections of ²³⁸U,²³⁸Pu, ²³⁹Pu, and those to decrease in k_eff are decreases in ν of ²³⁹Pu and increases in the capture cross sections of²⁴¹Am. They compensate each other, and the difference in k_eff between JENDL-3.3 and JENDL-4.0 is less than 0.1%Δk/kk′ and relatively small.     

Analysis of VIP-BWR reactor core physics experiments on UO2 and MOX mockup fuel assemblies - part 2: analysis with SIMULATE5

Analysis of the three test cores in the VIP-BWR program was performed in a two-dimensional geometrical model with CASMO5 coupled with the JENDL-4.0-based neutron data library, and reported in the previous paper. Following the study, interpretation of the experiments were carried out in a three-dimensional geometrical model with SIMULATE5 for the code validation study. The nuclear libraries for the SIMULATE5 calculations were generated with CASMO5 with the JENDL-4.0-based neutron data library. The effective multiplication factors of the critical cores ranged from 0.9983 to 1.0023 with measurement uncertainties of 0.0003 to 0.0004 (one σ). The root mean squares of (the calculated/the measured-1) for the fission rates at the core-mid plain of all the measured fuel rods were about 3% for the three cores. It was noticed that the calculations underestimated the fission rates of the UO₂ fuel rods and overestimated those of the MOX fuel rods for the test cores loaded with MOX fuel rods, which was consistent with trends in the preceding analysis studies of the VIP-BWR program and other MOX core experiments, and the biases were confirmed in the calculation results of power distributions in MOX-fueled light water reactor cores.     

An overview of instrumentation and control systems of a Korea standard nuclear power plant: A signal interface standpoint

This paper presents an overview of instrumentation and control (I&C) systems of a pressurized water reactor (PWR) type nuclear power plant (NPP) in Korea. Yonggwang unit 3, which was constructed as a basis model for a Korea standard nuclear power plant (KSNP), is selected as an example for the presentation. This overview is derived from analyzing the I&C systems based on a top-down approach. The I&C systems consist of 30 systems. The 183 I&C cabinets are also analyzed and mapped to the systems. The overview is focused on an interface between the systems and the cabinets. This information will be used to understand the implementation of the I&C systems and to group the systems for an upgrade.