Spectrometric method of evaluating spent nuclear fuel characteristics

A spectrometric method of identifying spent fuel assemblies according to the type of fuel elements present in them is described. The method is based on the results of spectrometric measurements and subsequent analysis of the radiation from fission products and the characteristic radiation from uranium in the irradiated fuel. The fuel assemblies used in the VVR-2 and OR research reactors contained fuel elements of a different type, differing by the initial quantity of uranium contained in them. To prepare the spent fuel assemblies for shipment to a reprocessing facility after long-time storage in cool-down pools, the assemblies must be sorted according to the type of fuel elements present in them. The method developed for identifying the types of fuel elements in the irradiated fuel is based on the dependence of the intensity of the characteristic radiation from uranium on the uranium content in a fuel element. The degree of excitation of the characteristic radiation of uranium also depends on the intensity of the radiation from fission products, which is monitored during the spectrometric measurements performed on the irradiated fuel; ultimately, this makes it possible to sort the spent fuel assemblies.     

Taking account of the effect of the curvature of VVÉR-1000 fuel assemblies on the power of the fuel elements

A comparative analysis is made of the deterministic and statistical methods of taking into account the effect of the curvature of VVéR-1000 fuel assemblies on the power of fuel elements. The fuel-element distribution of the energy release in the core for any random distribution of the gaps between the fuel assemblies is simulated, using the MEX code, on the basis of precise calculations (MCU code) and design calculations (BIPR-7 and PERMAK codes). The Monte Carlo method (Zazor code) was used to model the nominal density distribution of gaps in the core for different degrees of curvature of the fuel assemblies. It is shown that the power gain, obtained for the fuel elements by the probabilistic-statistical method, due to the curvature of the fuel assemblies is smaller and makes it possible to substantiate core safety with large perturbations, in contrast to the deterministic “maximum gaps near the most-energy stressed fuel element” method. 5 figures, 1 table, 3 references. Special Design Office “Gidropress.” Translated from Atomnaya énergiya, Vol. 87, No. 3, pp. 210–213, September, 1999.     

Method for detecting unsealed fuel elements in VVER and RBMK fuel assemblies and test results

A method is described for detecting unsealed fuel elements in VVER and RBMK fuel assemblies in a cooling pond. The method is based on detecting water which has seeped under the cladding of an unsealed fuel element. The results of testing the method on unsealed VVER-440, -1000, and RBMK-1000 fuel assemblies are presented. Translated from Atomnaya énergiya, Vol. 106, No. 2, pp. 84–88, February, 2009.     

Burnup studies of spent fuels of varying types and enrichment

This paper describes the results of fuel burnup measurements, made over a period of several years on discharged fuel from nuclear power plant and research reactor. The measured and calculated burnup of different spent fuel types, viz.: Natural uranium CANDU fuel bundles; 10.5% enriched booster rods; 20% enriched MTR fuel elements have been presented. High-resolution gamma spectrometry, using ¹³⁷Cs and ¹³⁴Cs burnup monitors was employed in different reactors to estimate the amount of ²³⁵U depletion in the respective fuel. The experimental data was compared with those of calculations to optimize fuel-scheduling programme. The burnup data is useful for assessment of fuel performance in the core and resolving design issues related to long-term storage facilities. It has been observed that the gamma spectrometry is very effective in identifying exact position of individual booster bundles inside the discharged booster assemblies, which is useful in safeguard applications. It is concluded that the distribution of measured isotopic activity ratios of ¹³⁴Cs/¹³⁷Cs along the height of the spent fuel gives accurate estimate of the axial neutron flux profiles in the core. The activity ratio technique therefore provides a useful method to determine flux peaking factors at the particular locations of the fuel assemblies in the reactor.     

Determination of the Fuel-Assembly Burnup in a Research Reactor by Repeated Short-Time Irradiation Followed by γ Spectrometric Measurements

Experiments performed to determine the absolute fuel burnup in spent fuel assemblies in the IRT research reactor at the Moscow Engineering Physics Institute are described. The method is based on measuring the residual amount of ²³⁵U in the spent fuel asemblies with respect to the activity of the fission product ¹⁴⁰La accumulated in fresh and spent fuel assemblies after they were irradiated for a short time in the reactor core. A fresh fuel assembly with known uranium mass was used as a standard. The neutron flux was monitored using Al + Cu and Al + Co wires placed at the center of the fuel assembly. Small corrections for the difference in the spectrum amd the flux density of the neutrons in fuel assemblies with different uranium content were obtained from the calculations. The burnup of the three fuel assemblies studied was determined to within less than 2%.     

Load characteristic of charged neutron detectors with a hydrogen-containing scatterer

Conclusions Our investigations showed that the double-cut method is suitable for mechanical polarization of fuel assemblies. The investigations made it possible to develop turnkey industrial equipment for cutting spent fuel assemblies, having different geometries, with a maximum size of up to 170 mm. The cutting unit is operating successfully at the Industrial Association “Mayak.” All-Union Scientific Research Institute of Thermophysical Apparatus. Translated from Atomnaya énergiya, Vol. 77, No. 3, pp. 194–203, September, 1994.     

Optimization studies of fuel loading pattern for a typical Pressurized Water Reactor (PWR) using particle swarm method

A three dimensional multi-energy group computer model PRISHA, which solves the neutron diffusion equations using finite difference method is developed for Pressurized Water Reactor (PWR). This computer code can find an optimum loading of a group of fresh fuel assemblies along with fuel assemblies of different exposures. The successive line over relaxation (SLOR) method is used to solve neutron diffusion equations. After validation of this part of computer code against an IAEA – PWR benchmark problem with 177 fuel assemblies in the core, particle swarm optimization (PSO) method is incorporated in the code for finding the optimum fuel loading pattern. A typical PWR core with 157 fuel assemblies, where 289 fuel pins are arranged in 17 × 17 rectangular arrays in a fuel assembly, was analyzed using this computer model for two cycles using PSO method. Different numbers of particles and iterations were used in PSO method. The results are found to be not very sensitive to either the number of particles or the number of iterations used in PSO method for considered case. However, a number of experiments have to be performed to arrive at the best global fitness parameter. Reasonably low power peaking factors were obtained for both the cycles.     

Description of the German Concept of Final Storage Cask

Abstract

The German storage concept for the direct final storage of spent fuel assemblies from LWR reactors is described. The final storage concept is designed in such a way that it encompasses the whole spectrum of fuel elements to be stored from German reactors, Le. U fuel assemblies and MOX fuel assemblies with a mean bumup of 55 GW.d.t^?1 heavy metal were considered. The further design requirements are defined in such a way that the cask concept satisfies the conditions for type B(U) transport, interim storage and fmal storage. The safe long-term containment of the activity is guaranteed by an inner cask welded leak-tight; the sufficient shielding and the transport packaging are ensured by a shielding cask.     

Assessment of passive autocatalytic recombiners' capabilities in case of Zr oxidation during station blackout scenario for TVSA and TVSM fuel assemblies of WWER 1000

This paper presents the investigation of Passive Autocatalytic Recombiners' (PARs) capabilities for hydrogen recombination in case of Station blackout scenario. The assessment was performed for both types of WWER fuel assemblies – the old Modernistic type of fuel assemblies (TVSM) and recently installed new Alternative type of fuel assemblies (TVSA) in Kozloduy NPP. The main difference between both types of fuel assemblies is the different geometries, masses of internals materials as well as different burnable poisons. To investigate the PARs' capabilities it has been performed comparison of fuel behaviour of both types of fuel assemblies.To perform this analysis it has been used MELCOR “input model” for Kozloduy Nuclear Power Plant (KNPP) WWER-1000. The model was developed at the Institute for Nuclear Research and Nuclear Energy (INRNE) for investigation of severe accident scenarios. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides' release outside fuel cladding.To assess the PARs' capabilities it was used the acceptance criterion for containment integrity to be 8% hydrogen concentration. This criterion was based on the decision of RSK (Germany commission for reactor safety).Generally, based on the performed analyses it was made a conclusion that using both types of fuel assemblies it was not disturbance of PARs' capabilities and safety criterion of NPP.     

快堆燃料组件少棒替代方法研究

快堆堆芯流量分配实验需大量燃料组件,为缩短燃料组件的加工周期,需寻找一种可简化燃料组件结构的思路。本文采用CTS理论算法,计算了燃料组件结构参数变化对组件水力特性的影响,提出了采用较少燃料棒替代组件完成试验的思路。该方法不改变组件外部结构与试验环境,仅用少量燃料棒获得与多棒燃料组件相同的水力特性。计算结果表明,替代组件与原组件水力曲线吻合较好,可达到替代效果。     

Application of gamma-neutron ionization chambers for monitoring the distribution of the neutron flux density,energy release,and energy production in RBMK-1000

A method developed for performing direct measurements of three-dimensional distributions of energy release and energy production in RBMK fuel assemblies is described. The method is based on performing measurements with a gamma-neutron chamber and comparing the neutron and gamma signals. The results of the measurements of the neutron flux density, energy release, and energy production are compared with the values obtained with the Prizma-M program of the Skala-micro information-measurement system. It is confirmed experimentally that the Prizma-M system can be used to monitor the distribution of not only the neutron flux density and energy release of fuel assemblies but also the energy production of off-loaded fuel assemblies. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 182–186, September, 2007.     

压水堆燃料辐照后燃料棒棒间距数据机器视觉测量方法研究

针对辐照后燃料棒棒间距数据获取和处理困难的问题,基于燃料棒几何特性及其在压水堆燃料组件中的排列方式,本文提出一种基于机器视觉的高效、可靠的燃料棒棒间距数据测量方法。该方法首先采用Retinex算法对水下燃料棒的采集图像进行增强预处理;然后,针对燃料棒阵列的前后成像干扰问题,采取边缘增强和逐行灰度特征处理方法实现待测燃料棒与背景燃料棒的有效分离,并进一步提升图像清晰度;最后,对燃料棒图像的单行灰度值进行二次曲线拟合,获得各个燃料棒的亚像素边缘点坐标。乏燃料组件的现场实验验证结果表明,该方法可一次性实现16个燃料棒棒间距测量,且测量精度达±0.32 mm,可为燃料性能分析提供高效、可靠的数据支持。      

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis.     

先进非能动压水堆核电站燃料组件轴向燃耗分布研究

燃料组件的轴向燃耗分布以及末端效应是燃耗信任制技术应用中的难点。基于先进非能动压水堆核电站的运行模式及组件设计特点,结合可能的燃料管理策略,统计轴向两端使用低富集度抑制区的乏燃料组件,生成轴向燃耗包络线。以 AP1000堆型的乏燃料贮存单元为例,通过分析统计,证明生成的轴向燃耗包络线用于临界安全分析是保守的。在此基础上,详细研究燃料组件顶部的低富集度抑制区对末端效应的贡献,并对燃料组件进行设计改进,减小至消除末端效应,为简化乏燃料组件相关的临界安全分析提供了一个方法。相关研究工作及成果,是先进非能动压水堆核电站乏燃料组件相关的设施设备的临界安全设计的基础,可为其他堆型的相应研究提供参考和借鉴。     

先进非能动压水堆核电站燃料组件轴向燃耗分布研究

燃料组件的轴向燃耗分布以及末端效应是燃耗信任制技术应用中的难点。基于先进非能动压水堆核电站的运行模式及组件设计特点,结合可能的燃料管理策略,统计轴向两端使用低富集度抑制区的乏燃料组件,生成轴向燃耗包络线。以AP1000堆型的乏燃料贮存单元为例,通过分析统计,证明生成的轴向燃耗包络线用于临界安全分析是保守的。在此基础上,详细研究燃料组件顶部的低富集度抑制区对末端效应的贡献,并对燃料组件进行设计改进,减小至消除末端效应,为简化乏燃料组件相关的临界安全分析提供了一个方法。相关研究工作及成果,是先进非能动压水堆核电站乏燃料组件相关的设施设备的临界安全设计的基础,可为其他堆型的相应研究提供参考和借鉴。     

Probabilistic Method for Estimating the Effect of the Gaps between Fuel Assemblies on the Energy Release in a VVéR-1000 Core

A method is presented for analyzing the influence of perturbations and uncertainties in the core on the energy release of fuel elements. The method uses a system of programs – design, precision, and specially developed programs – and it uses as initial data the results of complicated thermomechanical calculations of the curvature of fuel assemblies. A criterion is formulated for the usefulness of fuel-assembly profiling. The method as a computational tool makes it possible to develop improved profiling schemes which give higher heat-engineering margins of safety.     

Validation of nuclear and radiation safety of a container for spent amb reactor fuel assemblies at the Beloyarskaya nuclear power plant

An approach is proposed for validating the nuclear and radiation safety of a container for spent fuel assemblies from AMB-100 and-200 reactors at the Beloyarskaya nuclear power plant. To validate the radiation safety, the characteristics of fuel assemblies and their classification according to the average fuel burnup in the casing, and the intensities of n and γ radiation in the casing are analyzed. Nuclear safety is validated on the basis of the concept of a “model” casing. This model makes it possible to obtain an upper estimate of the effective coefficient of neutron multiplication for all real casings with fuel assemblies. Calculations are used to determine the minimum necessary thickness of the vessel, bottom, and cover for 17-and 35-place casings. It is shown that no special neutron protection is needed. The container design to be developed meets the IAEA and OPBZ-83 safety standards. __________ Translated from Atomnaya énergiya, Vol. 100, No. 6, pp. 423–428, June, 2006.     

(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.     

Comparison of hydrogen generation for TVSM and TVSA fuel assemblies for water water energy reactor (VVER)-1000

This paper presents the results received during investigation of hydrogen generation for both types fuel assemblies—the old modernistic type of fuel assemblies (TVSM) and recently installed new one alternative type of fuel assemblies (TVSA) in case of severe accident. There are some differences between both types FAs. They have different geometry as well as different burnable poisons. To investigate behavior of new fuel assemblies during the severe conditions it have been performed comparison of fuel behavior of old type TVSM fuel assembly to new one TVSA.To perform this investigation it has been used MELCOR “input model” for Kozloduy Nuclear Power Plant (KNPP) VVER 1000. The model was developed by Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) for investigation of severe accident scenarios and Probabilistic Safety Analyses (PSA) level 2. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides release outside fuel cladding.It was accepted criteria for vessel integrity about hydrogen concentration to be 8%. This criterion was based on the decision of RSK (Germany commission for reactor safety).Generally based on the received results it was made conclusion that using both types of fuel assemblies it was not disturbance safety conditions of NPP.     

Use of weapons plutonium in VVÉR-1000 reactors

One scenario for using excess Russian weapons plutonium is to load it into VVéR-1000 reactors. It is proposed that up to 40% of the fuel assemblies with uranium fuel be replaced with structurally similar fuel assemblies with mixed uranium-plutonium fuel. The stationary regime for burning fuel has the following characteristics: the run time is about 300 or 450 eff. days, the yearly plutonium consumption reaches 450 kg, the neutron-physical characteristics are close to the corresponding regimes with uranium fuel. The nuclear safety criteria and the irradiation dose for workers handling fresh and spent mixed fuel remain within the limits of the normative values. The use of mixed fuel makes it necessary to upgrade certain systems at nuclear power plants. A substantial quantity of weapons plutonium can be loaded every year into VVéR-1000 reactors, effectively using the energy potential of this plutonium. __________ Translated from Atomnaya énergiya, Vol. 103, No. 4, pp. 215–222, October, 2007.