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.

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.

Conceptual design of irradiation device for silicon neutron transmutation doping around Es-Salam research reactor

Silicon neutron transmutation doping remains one of the most viable nuclear applications for research reactors.Providing this kind of product involves an irradiation method capable of fulfilling the quality requirements of doping and alleviating the challenges related to the design and safety of the irradiation device.In this paper,we propose an irradiation device prototype for neutron transmutation doping of silicon ingots with diameters of 2 to 3 in.based on the Es-Salam research reactor.The thermal hydraulic analysis of the proposed irradiation device was performed to determine the optimum conditions for cooling.The effect of the mechanical vibrations induced by the circulation of coolant in the device was quantified via experimental measurements under different flow rates.The results show that the maximum temperature reached by the silicon ingots is below the temperature limit,effectively validating the design of the irradiation device.Other investigations are prospected to further optimize the design and the irradiation conditions.The irradiation of silicon ingots with a large diameter will be considered.

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.

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.

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.

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.

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.

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.

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.

Experimental and numerical studies of Ca(OH)_(2)/CaO dehydration process in a fixed-bed reactor for thermochemical energy storage

The Ca(OH)_(2)/CaO thermochemical energy storage(TCES)system based on calcium looping has received extensive attention owing to its high energy storage density,prolonged energy storage time,and environmental friendliness.The heat storage process of the Ca(OH)_(2)/CaO TCES system in a mixed heating reactor was evaluated in this study,by employing a combination of direct and indirect heating modes.The dehydration process was studied experimentally,and a numerical model was established and verified based on the experimental results.The dehydration behavior of 500 g of Ca(OH)_(2) powder was investigated in a fixed-bed reactor with mixed heating.The experimental and simulation results indicated that mixed heating causes combined centripetal and horizontal propulsion.Heat input is the main limiting factor in the heat storage process,because the radial advance of the reaction is hindered by the low thermal conductivity of the solid reactant particles.Heat transmission partitions were added to enhance the performance of the reactor.The performance of the modified reactor was compared with that of a conventional reactor.The radial heat transmission partitions in the modified reactor effectively enhance the energy storage rate and reduce the reaction time by 59.5%compared with the reactor without partitions.

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.

核电站堆腔混凝土辐照试验研究

作为核电站关键材料的堆腔混凝土,其安全服役是核电站长期稳定安全运行的前提条件之一。为了进行堆腔混凝土的中子辐照损伤机理研究,获得堆腔混凝土在中子辐照环境下的试验数据,本文建立了堆腔混凝土辐照试验方法,研制了辐照试验装置,并在研究堆中对其进行了加速辐照试验。结果表明:辐照试验装置设计合理,辐照试验指标满足试验要求,实现了两种规格多个混凝土试样的中子辐照。进一步的混凝土试样辐照性能研究结果表明:混凝土试样在平均快中子注量3.41×10^(18) cm^(−2)下辐照后,与辐照前相比,其外部形状未见明显差异,但试样颜色变化较大,并且出现一定的辐照肿胀和力学性能退化现象。

清华大学核研院研制5 MW低温核供热试验堆与10 MW高温气冷实验堆的工程技术创新

清华大学核能与新能源技术研究院(简称核研院)先后在1989年和2000年建成了5 MW低温核供热试验堆与10 MW高温气冷实验堆。在建堆过程中,清华大学核研院坚持设计创新与工具创新、工艺创新、工序创新密切结合,完成了一系列关键设备和零部件的制造与安装,使得整个工程项目顺利完工。在工程史研究中,技术工人做出的创新贡献并未引起学术界足够重视。本文表明,技术工人在工具、工艺、工序、制造与安装阶段的技术创新,亦是工程创新的重要保证。

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。