Conceptual design and safety characteristics of a new multi-mission high flux research reactor

Research reactors with neutron fluxes higher than 10^(14) n cm^(−2) s^(−1) are widely used in nuclear fuel and material irradiation,neutron-based scientific research,and medical and industrial isotope production.Such high flux research reactors are not only important scientific research facilities for the development of nuclear energy but also represent the national comprehensive technical capability.China has several high flux research reactors that do not satisfy the requirements of nuclear energy development.A high flux research reactor has the following features:a compact core arrangement,high power density,plate-type fuel elements,a short refueling cycle,and high coolant velocity in the core.These characteristics make it difficult to simultaneously realize high neutron flux and optimal safety margin.A new multi-mission high flux research reactor was designed by the Institute of Nuclear and New Energy Technology at Tsinghua University in China;the reactor can simul-taneously realize an average neutron flux higher than 2.0×10^(15) n cm^(−2) s^(−1) and fulfill the current safety criterion.This high flux research reactor features advanced design concepts and has sufficient safety margins according to the preliminary safety analysis.Based on the analysis of the station blackout accident,loss of coolant accident,and reactivity accident of a single-control drum rotating out accidently,the maximum temperature of the cladding surface,minimum departure from nucleate boiling ratio,and temperature difference to the onset of nucleate boiling temperature satisfy the design limits.

Studies on Production Planning of Dispersion Type U3Si2-Al Fuel in Plate-Type Fuel Elements for Nuclear Research Reactors

Several fuel plants that supply nuclear research reactors need to increase their production capacity in order to meet the growing demand for this kind of nuclear fuel. After the enlargement of the production capacity of such plants, there will be the need of managing the new production level. That level is usually the industrial one, which poses challenges to the managerial staff. Such challenges come from the fact that several of those plants operate today on a laboratorial basis and do not carry inventory. The change to the industrial production pace asks for new actions regarding planning and control. The production process based on the hydrolysis of UF6 is not a frequent production route for nuclear fuel. Production planning and control of the industrial level of fuel production on that production route is a new field of studies. The approach of the paper consists in the creation of a mathematical linear model for minimization of costs. We also carried out a sensitivity analysis of the model. The results help in minimizing costs in different production schemes and show the need of inventory. The mathematical model is dynamic, so that it issues better results if performed monthly. The management team will therefore have a clearer view of the costs and of the new, necessary production and inventory levels.

Calculation of dpa rate in graphite box of Tehran Research Reactor(TRR)

Radiation damage is an important factor that must be considered while designing nuclear facilities and nuclear materials. In this study, radiation damage is investigated in graphite, which is used as a neutron reflector in the Tehran Research Reactor (TRR) core. Radiation damage is shown by displacement per atom (dpa) unit. A cross section of the material was created by using the SPECOMP code. The concentration of impurities present in the non-irradiated graphite was measured by using the ICP-AES method. In the present study the MCNPX code had identified the most sensitive location for radiation damage inside the reactor core. Subsequently, the radiation damage (spectral-averaged dpa values) in the aforementioned location was calculated by using the SPECTER, SRIM Monte Carlo codes, and Norgett, Robinson and Torrens (NRT) model. The results of “Ion Distribution and Quick Calculation of Damage”(QD) method groups had a minor difference with the results of the SPECTER code and NRT model. The maximum radiation damage rate calculated for the graphite present in the TRR core was 1.567 9 10^-8 dpa/s. Finally, hydrogen retention was calculated as a function of the irradiation time.

Neutronic design investigation of a liquid injection-based second shutdown system for a typical research reactor using MCNPX

Safety systems, built on state-of-the-art technology, are essential for achieving acceptable levels of plant safety to minimize hazards to the reactor and the general public. The second shutdown system(SSS) as an engineered safety feature and a part of the reactor protection system(RPS) is a means for rapidly shutting down a nuclear reactor, keeping it in a subcritical state and serving as a backup to the first shutdown system(FSS). In this research, one SSS with two types of optimum chamber designs is proposed that take into account the main current characteristic features of the Tehran research reactor with improvements over earlier designs. They are based on a liquid neutron absorber injection that is preferably different, diverse, and independent from the FSS based on the rod drop mechanism. The major design characteristics of this SSS with two different chambers were investigated using MCNPX 2.6.0 code. The performed calculations showed that the designed SSS is a reliable shutdown system, assuring an appropriate shutdown margin and injection time, with no significant effects on the effective delayed neutron fraction while causing minimal variations to the core structure. Further, the reasonable financial cost and the prolongation of the operation cycle are additional advantages of this design.

Effective point kinetic parameters calculation in Tehran research reactor using deterministic and probabilistic methods

The exact calculation of point kinetic parameters is very important in nuclear reactor safety assessment, and most sophisticated safety codes such as RELAP5, PARCS,DYN3D, and PARET are using these parameters in their dynamic models. These parameters include effective delayed neutron fractions as well as mean generation time.These parameters are adjoint-weighted, and adjoint flux is employed as a weighting function in their evaluation.Adjoint flux calculation is an easy task for most of deterministic codes, but its evaluation is cumbersome for Monte Carlo codes. However, in recent years, some sophisticated techniques have been proposed for Monte Carlo-based point kinetic parameters calculation without any need of adjoint flux. The most straightforward scheme is known as the ‘‘prompt method'' and has been used widely in literature. The main objective of this article is dedicated to point kinetic parameters calculation in Tehran research reactor(TRR) using deterministic as well as probabilistic techniques. WIMS-D5B and CITATION codes have been used in deterministic calculation of forward and adjoint fluxes in the TRR core. On the other hand, the MCNP Monte Carlo code has been employed in the ‘‘prompt method''scheme for effective delayed neutron fraction evaluation.Deterministic results have been cross-checked with probabilistic ones and validated with SAR and experimental data. In comparison with experimental results, the relativedifferences of deterministic as well as probabilistic methods are 7.6 and 3.2%, respectively. These quantities are10.7 and 6.4%, respectively, in comparison with SAR report.

Neutronic Analysis of Generic Heavy Water Research Reactor Core Parameters to Use Standard Hydride Fuel

This research presents neutronic calculations for heavy water research reactor core substituting hydride fuel for uranium dioxide fuel. The aim of this research is feasibility analysis of reactor utilization with its original design using a new proposed fuel and changing the coolant and moderator circuit to light water. The required group constants for the CITATION code will be calculated using WIMSD-4 code. Neutronic calculations such as multiplication factors, radial and axial power peaking factor and fuel burn-up calculations are carried out by the CITATION code.

Determination of fuel burnup distribution of a research reactor based on measurements at subcritical conditions

This paper presents the determination of the fuel burnup distribution of the Dalat nuclear research reactor(DNRR) using a method of measurements at subcritical conditions. The method is based on the assumption of linear dependence of the reactivity on the burnup of fuel bundles and the measurements at subcritical conditions.The measurements were taken for seven selected fuel bundles in two different measuring sequences. The measured burnup values have also been compared with the calculations for verifying the method and the measurement procedure. The results obtained with the three detectors have a good agreement with each other with a discrepancy less than 1.0%. The errors of the measured burnup values are within 6%. Comparison between the calculated and measured burnup values shows that the discrepancy of the C/E ratio is within 9% compared to unity. The results indicate that the method of measurements at subcritical conditions could be well applied to determine the relative burnup distribution of the DNRR.

Effects of cooling channel blockage on fuel plate temperature in Tehran Research Reactor

In this study, the variation of the temperature distribution of the fuel plate in Tehran Research Reactor core was studied in case of coolant channels blockage. While the experimental method is not possible, both the analytical and simulation methods were used to obtain the more reliable data. The results show that one channel blockage will increase the fuel temperature to about 100%, but it does not lead to clad melt down still. With further calculation and simulation it is understood that if the coolant velocity drops to 90% of its nominal value, it may causes the clad melting down. At least two channels with complete blockage even at the positions far from the core center can also melt down the clad.

Current status and technology development tendency of research reactors in China

The current status and development history of domestic and abroad research reactors (RRs) are mentioned.The representative RRs and their respective technology characteristics are introduced.The utilizations of China's RRs,mainly included as nuclear engineering technology,basic research applications of nuclear technology,teaching and personnel training,are explained.

Neutron Flux Signal Acquisition from Plant Instrumentation Channel of Research Reactor for Reactivity Calculation

A design for instantaneous neutron flux signal acquisition system is being carried out for reactivity measurement of the nuclear research reactor. It is a computer-based digital data acquisition system that can perform continuous monitor and measurement of reactivity inserted into or removed from the research reactor. The acquisition system accomplishes with two major parts. The first part is an interfacing PCI based data acquisition card and the corresponding driver software intending to on-line acquisition of neutron flux signals from plant instrumentation channel. The second part incorporates the high-level Visual Basic real time program, indigenously developed for computation of reactivity by the solution of neutron point kinetic equations and other relevant functional modules like input file logging, reactivity calculation, graphics demonstration etc.

Measurement of Natural and Artificial Radioactivity in Soil at Some Selected Thanas around the TRIGA Mark-II Research Reactor at AERE, Savar, Dhaka

The activity concentration of natural and fallout radionuclides in the soil at some selected Thanas around the TRIGA Mark-II Research Reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka were measured by using a high purity germanium detector (HPGe). The study revealed that only natural radionuclides were present in the samples and no trace of any artificial radionuclide was found. The average activity concentration of 238U, 232Th and 40K were found to be 37.8 ± 5.6 Bq.kg-1, 58.2 ± 11.0 Bq.kg-1 and 790.8 ± 153.4 Bq.kg-1 respectively. The radium equivalent activity (Req), absorbed dose rate (D), external radiation hazard index (Hex) and internal radiation hazard index (Hin) were also calculated to find out the probable radiological hazard of the natural radioactivity.

Development of 24 and 59 keV Filtered Neutron Beams for Neutron Capture Experiments at Dalat Research Reactor

External filtered neutron beams have been developed at the horizontal radial channels No. 4 of Dalat research reactor. In the material composition of the neutron filters, the primary material components of Iron, Aluminum, Nickel and Vanadium are used to obtain the mono-energetic neutron beams of 24 and 59 keV, with low level of Gamma and slow neutron background. A computer code and Monte-Carlo simulation technique were applied to optimize the filter configurations and to deduce the neutron energy distributions in the filtered beams. A hydrogen-filled proton recoil detector and the activation method with Gold foils were used to measure the neutron energy spectrum and flux of each beam at sample position. The results of experimental neutron fluxes are 6.1 × 105 and 5.3 × 105 n/cm2/s for 24 and 59 keV beams, respectively.

Thermal Hydraulic Analysis Improvement for the IEA-R1 Research Reactor and Fuel Assembly Design Modification

This paper presents the sequence of activities to improve the thermal hydraulic analysis of the IEA-R1 research reactor to operate in safe conditions after power upgrade from 2 to 5 MW and core size reduction from 30 to 24 fuel assemblies. A realistic analysis needs the knowledge of the actual operation conditions (heat flow, flow rates) beyond the geometric data and the uncertainties associated with manufacturing and measures. A dummy fuel assembly was designed and constructed to measure the actual flow rate through the core fuel assemblies and its pressure drop. First results showed that the flow distribution over the core is nearly uniform. Nevertheless, the values are below than the calculated ones and the core bypass flow rate is greater than those estimated previously. Based on this, several activities were performed to identify and reduce the bypass flow, such as reduction of the flow rate through the sample irradiators, closing some unnecessary secondary holes on the matrix plate, improvement in the primary flow rate system and better fit of the core components on the matrix plate. A sub-aquatic visual system was used as an important tool to detect some bypass flow path. After these modifications, the fuel assemblies flow rate increased about 13%. Additional tests using the dummy fuel assembly were carried out to measure the internal flow distribution among the rectangular channels. The results showed that the flow rate through the outer channels is 10% - 15% lower than the internal ones. The flow rate in the channel formed between two adjacent fuel assemblies is an estimated parameter and it is difficult to measure because this is an open channel. A new thermal hydraulic analysis of the outermost plates of the fuel assemblies takes into account all this information. Then, a fuel design modification was proposed with the reduction of 50% in the uranium quantity in the outermost fuel plates. In order to avoid the oxidation of the outermost plates by high temperature, low flow rate, a reduction of 50% in the uranium density in the same ones was shown to be adequate to solve the problem.

Studies on Capacity Expansion of Fuel Plants for Nuclear Research Reactors

The demand for nuclear fuel for research reactors is rising worldwide. Thus, the production facilities of this kind of fuel need reliable guidance on how to augment their production in order to meet the increasing demand efficiently and safely. We proposed a specific procedure for increasing production capacity. That procedure was tested with data from a real plant, which produces plate-type fuel elements loaded with LEU U3Si2-Al fuel. The test was made by means of discrete event simulation, and the results indicated the proposed procedure is efficient in raising production capacity.

Application of silicon carbide temperature monitors in 49-2 swimming-pool test reactor

High purity SiC crystal was used as a passive monitor to measure neutron irradiation temperature in the 49-2 research reactor.The SiC monitors were irradiated with fast neutrons at elevated temperatures to 3.2×10^(20)n/cm^(2).The isochronal and isothermal annealing behaviors of the irradiated SiC were investigated by x-ray diffraction and four-point probe techniques.Invisible point defects and defect clusters are found to be the dominating defect types in the neutron-irradiated SiC.The amount of defect recovery in SiC reaches a maximum value after isothermal annealing for 30 min.Based on the annealing temperature dependences of both lattice swelling and material resistivity,the irradiation temperature of the SiC monitors is determined to be~410℃,which is much higher than the thermocouple temperature of 275℃ recorded during neutron irradiation.The possible reasons for the difference are carefully discussed.

中子成像技术在元素分析中的应用

中子成像作为一种快速、直观的无损检测技术,在核工业、航空航天、新能源、地质、考古、先进制造等多个领域得到广泛应用。中子成像利用中子不带电、穿透能力强、对轻元素敏感、可区分同位素和近邻元素等特性,非常适合开展含氢元素、近邻元素和同位素等材料的无损检测。通过概述中子成像技术的基本原理及特点,并结合中国先进研究堆(CARR)中子成像装置上的应用案例,重点综述了国内外中子成像技术在储氢材料、燃料电池、岩石、核燃料元件、古代文物等领域的典型应用。随着中子成像技术的不断发展和广泛应用,有望为我国更多领域研究提供更强有力的技术支撑。

小角中子散射原位热力耦合加载装置

热力耦合近工况条件下材料微观结构的原位实验研究,对于深入理解材料服役性能演化机制十分重要,可给出样品微观上的纳米结构尺度分布。为充分发挥小角中子散射统计性好、取样体积大可开展原位实验等优势,本文基于中国先进研究堆小角中子散射谱仪,设计并研制了一台高温和拉力同时加载的原位实验装置,并实现了高温高压下原位测量材料的纳米尺度形貌变化。实验测试结果表明,装置最大载荷可达20 kN,最高温度800℃,控温精度优于±1℃。利用该装置对镍基单晶高温合金样品进行了原位小角中子散射测试,发现温度拉力条件下样品内部纳米结构的明显变化,表明基于该装置可开展热力耦合加载下的原位小角中子散射实验。该装置及其相应实验方法,可用于核电不锈钢等多种高温结构材料的原位加载实验研究,提供微观结构演化数据。

基于新型泳池式研究堆BNCT中子束流装置方案设计研究

本研究基于新型泳池式研究堆设计方案,开展了硼中子俘获治疗(BNCT)能谱可变中子束流装置的初步方案设计。根据新型泳池堆堆芯结构,采用屏蔽体及准直器组合方式,对BNCT的中子慢化层、热中子吸收层、伽马屏蔽层以及中子准直器进行了分析计算及优化,在不增加中子引出束流孔道数量的前提下,实现了超热中子及热中子BNCT束流的切换,通过理论计算分析确定了两种装置的中子束流特性,超热/热中子通量密度、单位快中子剂量、单位光子剂量、热中子通量占比等参数均符合IAEA-TECDOC-1223报告的BNCT推荐参考标准,可用于不同能量需求的硼中子俘获治疗,为新型多功能泳池堆的应用及推广提供了技术支持。

HFETR同位素靶件的焊接工艺优化

为了提高产品合格率与生产效益,依据研究堆医用同位素靶件的结构和材料特性,研究了较优的焊接参数与合理可行的焊接工艺技术路线,同时根据实际焊接质量对靶件结构进行改进和优化,满足了研究堆医用同位素靶件随堆辐照需求,保障了同位素生产的顺利进行。采用该焊接方法和焊接工艺已成功完成多批次医用同位素靶件的焊接、生产及入堆辐照。

CARR重水浓缩监控系统改进与实现

为了满足中国先进研究堆(CARR)运行深层次的功能要求,对原有重水浓缩监控系统进行了升级改进,软件界面进行了重新开发设计,提供了友好的人机操作界面,在原系统中增加了“电解电源功率调节”和“通风流量”的监测通道和远程操作员站,实现了重水浓缩生产工艺的自动化控制,提高了重水浓缩的生产效率,保证了重水浓缩系统的安全性和可靠性。另外,CARR重水浓缩监控系统的改进也可为其他反应堆类似系统的实现具有借鉴意义。