bluuebunny/crossref_metadata_2025 huggingface dataset:
Active Interrogation of Highly Enriched Uranium in the Suitcase by Using NG-9 Neutron Generator
Authors: Shi Guang-Yu , Jing Shi-Wei
Published: November 2022

Preventing the proliferation of highly enriched uranium (HEU) is an important issue all over the world. The active interrogation of HEU hided in the suitcase based on neutron technique is studied by using MCNP5 code. The neutron source of the detection system is based on the NG-9 neutron generator developed by Northeast Normal University. A set of HEU detection devices has been established. Lead is a kind of neutron-multiplier material with high density, so the lead block is chosen as the substitute for HEU. The size of the aluminum alloy suitcase is 58 cm × 42 cm × 25 cm. The NG-9 D-T neutron generator is placed on the suitcase. A lead block measuring 26.5 cm × 12 cm × 1 cm is placed in the center of the suitcase. The lead block has a density of 11.35 g/cm and a mass of approximately 3.6 kg. Daily clothes are placed inside the suitcase as a distraction. A cylindrical BGO detector with a diameter of 7.2 cm, a height of 29.7 cm is placed close to the suitcase to record gamma rays. A cylindrical lead shield with a thickness of 5 cm is placed outside the BGO detector. Paraffin wax is placed around the whole detection device to protect the neutron radiation and avoid the interference of other substances on the detection results. The purpose of this experiment is to verify the agreement of the MCNP5 simulation results and the experimental results. In this paper, threshold energy neutron analysis (TENA), fast neutron method, and thermal neutron method are used to detect HEU. The simulation results show that the presence of HEU in the suitcase can be determined by neutron flux which is higher than those in the absence of HEU.
Study on Vertical Transfer of Radionuclide Co-60 in Broun Soil and Aeolian Sandy Soil Around Nuclear Power Plant
Authors: Zhang Qiong , Yuanlv Ye , Bo Wang , Lu Chen
Published: November 2022

The coastal soil surrounding to the Nuclear Power Plants were mainly classified to brown soil and aeolian sandy soil in China. In previous research, focused on the elements of Cs-137 and Sr-85 vertical transfer in these soils, but the Co-60, which was the most important factor in the brown soil and aeolian sandy soil were still urgent research. In this study, experiments were conducted to determine the vertical transfer of Co-60 in brown soil and aeolian sandy soil. The soil around Shidaowan Nuclear Power Plant were studied by using undisturbed soil column methods and isotope tracer technology, and identify the factors influencing their migration depths in soil. The results show that with the increase of leaching test, the migration of Co-60 in aeolian sandy soil was greater than that in brown soil, but the difference gradually decreases with the progress of spray test; the content of Co-60 in leaching water was very few and cannot be detected due to the strong adsorption by soil; after three years of undisturbed soil column experiment, 72.36 ∼ 85.26% of Co-60 mainly was retarded in the range of 0∼5cm in both of the soil surface. It was proved that Co-60 was difficult to migrate to groundwater in a short period. The distribution of specific activity of Co-60 in soil was an individual exponential declining with depth of oil. The results showed that the radionuclide Co-60 had little impact on groundwater in the years after the nuclear accident, but it should pay attention to the change of radioactivity concentration in surface soil and then select appropriate soil remediation methods. The study will provide a scientific basis for post-accident environmental impact assessment and post-accident soil remediation in China.
A Novel Transmission Reconstruction Algorithm for Radioactive Drum Characterization
Authors: Yang Hui , Zhou Hao , Dong Bing , Zhou Wentao , Gu Weiguo , Zhang Xinyu , Lei Qingxin , Shan Chenyu , Wang Dezhong
Published: November 2022

The accuracy of tomographic gamma scanning transmission reconstruction is a critical factor in reconstructing the activity of a radioactive drum. Traditional reconstruction algorithms produce severe grid artifacts and a high level of noise, thereby increasing the reconstruction error for both the density map and the activity. This paper proposes a novel algorithm for transmission reconstruction by combining maximum-likelihood expectation maximization and a convolutional neural network (CNN). Our experimental results indicate that the proposed reconstruction algorithm is capable of significantly reducing measurement errors, increasing spatial resolution while also eliminating grid artifacts, and being sufficiently robust when dealing with a noisy input image. The mean squared error of the output image decreased by 52.70% compared with the conventional reconstruction method, and the peak signal-to-noise ratio and structural similarity index improved by 21.89% and 17.33%, respectively. The spatial resolution was improved by 28 times, which demonstrates that CNN is a potentially useful new method for radioactive waste drum transmission image reconstruction.
Research on Risk Management of Nuclear Engineering - Taking T Project Reverse Power Supply As an Example
Authors: Wang Baoquan
Published: November 2022

Under the vision of sustainable development and a community with a shared future for mankind, the era of “zero carbon” is coming to us. Nuclear energy has long been seen as one of the powerful means to achieve a “carbon neutral” future. Nuclear power projects are characterized by high technical requirements, long construction period, complex system composition, numerous stakeholders and heavy social responsibilities. Any link is crucial. Therefore, effective risk management of nuclear power projects is of great significance.

Based on the actual situation of nuclear POWER T project, the following studies are carried out:

(1) Through the research of domestic and foreign literature, Delphi method anonymous voting and expert consultation meeting were adopted to merge, filter and supplement the risk factors, and 24 risk factors in seven categories, including design, installation and commissioning, construction, contract, external, procurement and administrative licensing, were identified in the process of power transmission of NUCLEAR POWER T project.

(2) AHP fuzzy comprehensive evaluation method is used to select the hierarchical structure model of power supply risk factors, construct a judgment matrix, establish a risk evaluation table, calculate the risk weight of each level and rank it, compare the value interval of risk evaluation set, and get the power supply risk level of NUCLEAR power T project.

(3) In the outlook, the author considers the use of bow-tie model to prevent and control risk events in advance in the future, so as to formulate measures. Finally, the integrated risk control platform of reporting, early-warning, disposal and monitoring is proposed to form a closed-loop management of risk identification, reporting, disposal and rectification.
Application of the Refinement Spacer Grid Model in Subchannel Analysis Code
Authors: Dong Si Ying , Yang Ping , Gan Fujun , Liu Yuanrong
Published: November 2022

As one of the important components in PWR cores, spacer grids can maintain the position of the rods during operation as well as promote fluid motion. The overall physical phenomena and local effects introduced by spacer grids can be accounted for the crossflow mixing model. Based on the in-house thermal-hydraulics (T/H) subchannel code SAPRE, the four improvements are developed:1) The distributed resistance method (DRM) is applied to develop the crossflow mixing model in order to reflect mixing vanes grid’s contribution to crossflow mixing in rod bundles. 2) By taking more detailed grid parameters into account, the loss coefficient of grid is calculated based upon form and friction losses. 3) The Gunter-Shaw frictional relationship is established to simulate the transverse resistance caused by the spacer mixing vanes. 4)Improvement of lateral momentum equation. Due to the presence of the mixing vanes, the crossflow is increased, and the domination of axial flow in the lateral momentum equation is not appropriate. The momentum changes caused by the crossflow on the flow term should be considered. All these four models are implemented in the subchannel code SAPRE. The improved SAPRE code is validated through Hassan 5 × 5 Particle Image Velocimetry experiment and PSBT benchmark. Numerical results demonstrate that SAPRE has large advantages in accurate prediction of local T/H parameters in rod bundles, which is of significance to safety and economy of PWR.
Microstructural Evolution of SiC-Coated C/C Composites Under Argon Ion Irradiation
Authors: Xie Xiangmin , Yan Long , Cheng Guodong , Tang Xian
Published: November 2022

SiC coatings have been used to improve the oxidation resistance and stability of C/C composites in high-temperature reactors. However, the irradiation-induced surface structural transformations of SiC-coated C/C composites have been rarely studied. Herein, chemical vapor reaction (CVR) SiC-coated C/C composites were irradiated with 300 keV argon ions at room temperature with irradiation doses ranging from 5 × 1015–1 × 1017 ions·cm−2. The damage patterns of the pristine C/C composites and SiC-coated C/C composites were observed using scanning electron microscopy, and the shape and size evolutions of the CVR-SiC particles were investigated as a function of the irradiation dose. The results revealed that the pristine C/C composites were severely damaged after ion irradiation, and a large number of defects and pores formed on the surface. In contrast, the ion-irradiated SiC-coated C/C composites showed an undamaged surface. As the irradiation dose increased from 0 to 1 × 1017 ions·cm−2, the CVR-SiC particles were transformed from irregular to spherical shapes, and the average size of the SiC particles was reduced from 22 to 5 μm. The size reduction and spheroidization of the SiC particles under irradiation were attributed to the amorphous transformation of SiC. This study can provide deeper insight into the irradiation behavior of SiC-coated C/C composites in high-temperature reactors.
On the Creation of Nuclear Law as a New Inter-Discipline in Law
Authors: Liu Jiu
Published: November 2022

Civil nuclear industry has developed rapidly and made remarkable achievements all over the world since 1940s. The nuclear legal system plays an important role in promoting and guaranteeing the development of civil nuclear industry and utilization of nuclear energy. The implementation and improvement of nuclear legal system form the corresponding discipline of nuclear law. The construction of nuclear law as a new inter-discipline of both nuclear science and technology and law needs a timely follow-up. Therefore, this paper starts from the analysis of the nuclear law as a branch law and studies the nuclear law as a new inter-discipline in law. Nuclear laws is still underway and needs to be improved currently. In order to create the new inter-discipline of nuclear law, it is necessary to make clear the definition, object and characteristics of the discipline, and to design well in terms of teachers and curriculum so as to make the discipline of nuclear law more mature and realize its value as a new inter-discipline in law.
Research on the Design Method of Nuclear Heat Exchanger Tube Based on Multi-Objective Optimization
Authors: Cao G. M. , Dai L. , Zhao Z. X. , He T.
Published: November 2022

In addition to the traditional performance indexes such as heat exchange efficiency and flow resistance, the design of nuclear heat exchange tube needs to take into account the flow noise level index of heat exchange tube at the same time. Therefore, the design of heat exchange tube needs to consider multiple properties such as heat exchange efficiency, flow resistance and flow noise at the same time, which greatly increases the complexity of nuclear heat exchange tube design. In this paper, a design method of low-noise nuclear heat exchange tube is proposed. The design parameters of nuclear heat exchange tube are automatically adjusted based on the coupling of multiple simulation software. The heat exchange efficiency, flow resistance and flow noise of nuclear heat exchange tube are characterized by normalizing the comprehensive evaluation factor of flow and heat transfer and insertion loss, and the optimal design parameters of nuclear heat exchange tube are found through orthogonal test.
Criticality Safety Analysis of Fuel Storage Under Fuel Assembly Dropping Accident
Authors: Dai Mingliang , Fu Pengtao , Zhao Jun
Published: November 2022

During fuel storage, lifting fuel assembly is an essential process. The fuel assembly dropping accident may lead to consequences, such as fuel assembly lying on the storage rack, fuel assembly deformation leading to out of control of fuel rod pitch distance, fuel rod breakage and other accidents, and then affect the critical safety of fuel storage system. Therefore, it is necessary to analyze the critical safety of fuel assembly storage under fuel assembly dropping accident.

This study analyzes all kinds of accident conditions that may occur under fuel assembly dropping accident, and uses Monte Carlo particle transport code JMCT to simulate the neutron effective multiplication factor (keff) at typical conditions. The results show that the calculated keff of the fuel storage system will not exceed the criteria when the fuel assembly drops on the fuel storage rack without deformation. When fuel assembly is deformed due to dropping from a height, change of fuel rod pitch distance in fuel assembly will affect the moderator-uranium ratio of the fuel assembly storage system, which may lead to an increase in keff in the system. However, the fuel storage system can also will not exceed the criteria when 2000 ppm soluble boron acid is taken credit in the spent fuel pool. Based on the above analysis, the fuel storage system can ensure the subcritical state at different condition of fuel assembly dropping.
Study on the Water Intake of Heat Trap Blocked by Marine Life in Nuclear Power Plant
Authors: Zhang Zhenhua , Zhang Bo , Zheng Chaoying
Published: November 2022

In recent years, the operational events that the water intake of heat trap was blocked by marine life in nuclear power plants have occurred repeatedly. Since 2000, there have been more than 200 shutdown events of biological (foreign body) blockage resulting in failure of cooling water system in the United States, the United Kingdom, Sweden, Arabia, India, Israel, Japan, South Korea and China. It has caused major safety risks and brought huge economic losses and has attracted wide attention from various governments, related industries and research institutions. It can be seen that water intake blockage event is a safety problem for all kinds of power plants to face. This paper introduces the typical design of water intake and analyses the potential effects of a large number of marine life into the heat trap water intake from the perspective of circulating water system (CRF) jump pump and essential service water system (SEC) loss of function. Based on the characteristics of the high frequency of such operation events and they may cause major safety risks, the reactor shutdown response time in different conditions is simulated. When a large number of marine organisms flood into the water intake of heat trap, the response of the operator on the potential impact of the unit condition and the operational response plan are analyzed. The impact of the unplanned changes in operating status caused by the blockage of the water intake is explained. Ultimately, based on the analysis as described above, suggestions on the design improvement of the heat trap water intake and seawater filtration system in the nuclear power plant are presented. At the same time, suggestions are put forward on the supervision analysis, dynamic monitoring of the surrounding environment of the water intake, establishing a rapid decision-making mechanism, improving the cold source emergency response plan and increasing the operation control strategy.
Research on Optimization of Reload Startup Physics Tests in Sanmen Nuclear Power Plant
Authors: Shi Jianfeng , Yang Qingxiang , Wang Lihua , Chen Lijiang , Li Ang
Published: November 2022

During reload startup of a nuclear power plant (NPP), physics tests are performed to verify the consistency of core neutronic characteristics with design expectations. Physics tests time during power ascension directly affects the power rise rate and economy of units operation. The paper studied the physics tests process during power ascension of reload startup in Sanmen NPP, and proposed an optimization scheme for Sanmen reload startup physics tests based on test data analysis and simulation. On the premise of meeting the requirements of current guidelines and standards, the scheme optimizes the current test method from the aspects of test contents andtest initial conditions, which can effectively reduce risk to core operation, shorten test time and improve economy of units operation. In terms of test content, it was demonstrated that the influence of detector leakage current on the power distribution measurement can be ignored, which can reduce the times of leakage current measurement in the process of power rise. In terms of the test initial conditions, the influence of xenon on the test process and results was analyzed, so as to relax the initial conditions of xenon equilibrium and shorten the total power rise time from about 4 days to about 2 days. Standard compliance was also evaluated, and the optimizations are in accordance with the requirements in current standards.
Numerical Investigation on Transport Characteristics of Hydrogen-Air-Steam Mixture Induced by Condensation in the Single Compartment
Authors: Liu Feng , Sun Zhongning , Bian Haozhi , Ding Ming
Published: November 2022

Under the severe accident of the nuclear power plant, hydrogen-steam mixture gas will be released into the containment compartments. In this process, there is a strong coupling relationship between mixture gas diffusion and steam condensation on the compartment wall surface. Especially for local compartments with poor connectivity, steam condensation is likely to aggravate the risk of hydrogen accumulation and combustion. This coupling phenomenon, which has not been systematically studied in previous numerical simulations or experiments, is especially complex and needs to be further studied. Moreover, the applicability of mechanistic numerical models for predicting condensation-induced transient gas transport processes also should be further confirmed. In this study, the mechanistic numerical model was evaluated through the transient hydrogen mixing experiment of the THAI facility. Based on the validated model, the transient mass transfer process of the mixture gas induced by condensation in a typical single compartment during the steam-hydrogen release process is studied. Furthermore, the influence of release condition of gas source term, wall sub-cooling on the hydrogen accumulation phenomenon in the compartment was discussed. The conclusion is helpful to further understand the transport behavior of hydrogen-air-steam mixture gas in the local single compartment under the severe accident.
Simulation of MSLB+SBO Nuclear Accident Drill Sequence of 3rd-Generation Passive Nuclear Power Plant Based on Severe Accident Simulator
Authors: Yang Yongyong , Huang Xiong , Wei Wei , Tao Yinyong , Xie Zhengquan , Ma Guoyang
Published: November 2022

The full scope simulator (FSS) is very important for operator training, but the related SA training can’t be conducted for lack of the severe accident models, so it is urgent to expand the severe accident model on the FSS in nuclear power plant (NPP). The full scope simulator of 3rd-generation passive NPP can perform more abundant and complete accident scenes for operator training and accident drills through coupling MAAP5 models, namely severe accident simulator (SA simulator). In this paper, the severe models (such as core, primary system, containment), the key technologies of coupling MAAP5 software with FSS are described. Then taking the MSLB+SBO as the initial hypothetical condition to simulate the whole process of the severe accident emergency drill scenario in 3rd-generation NPP and make a preliminary sensitivity analysis on the mitigation effect of the mitigation measure for RCS injection flow in the accident drill scenario. The results show that the whole severe accident drill process can be very simulated, the SA simulator can not only provide an effective mean for the simulation of NPP severe accident drill scenarios, but also can be used as an auxiliary tool for the verification of severe accident management guidelines (SAMG) and the optimization of drill scenarios in NPP, which provide some technical support for the safe and stable operation of nuclear power plant.
Research on Quantitative Assessment Method of Security Guard Capacity in Nuclear Power Plants
Authors: Liu Xiaolin , Meng Wei , Li Jianwen , Zhang Jinfei
Published: November 2022

Under current circumstances, the daily training of the security team in nuclear power plant is carried out by the security company providing security services according to the requirements of the power plant security department, followed by manual assessment and authorization among the guards. There are several problems in this process that can be improved: evaluation is highly subjective, difficult to quantify, and difficult to standardize. Based on the principles of objectivity, unity, comprehensiveness and extendibility, this paper proposes an intelligent assessment method for nuclear power plant security guard capability. Firstly, it decomposes single task into core abilities which are aggregated to an ability library based on their corresponding tasks. Secondly, all task scenarios are made into VR test scenarios with core abilities as the test points, and aggregated into test banks. Then use the VR assessment system for automatically issuing, collecting and scoring test papers. Finally, a comprehensive assessment of the tested guard will be released with the combination of the fuzzy comprehensive assessment method. This method not only can increase objectivity and scientificity of assessment but also has extendibility to new job requirements to meet dynamic changes.
Optimization for Rapid Dose Calculation of the RIMPUFF Model and Its Evaluation Against Belgian Field Experiment
Authors: Yang Li , Li Xinpeng , Chen Yixue , Wang Boxin , Fang Sheng
Published: November 2022

The Risø Mesoscale PUFF model (RIMPUFF) is a Lagrangian atmospheric dispersion model that uses consecutive Gaussian puffs to simulate accidental release which is widely used in nuclear emergencies. Based on the real-time puff size and the distance between the puff and the monitoring station, RIMPUFF can calculate the dose results quickly by using the dose calculation interpolation table within its dose calculation module but the results are biased. To improve the accuracy of dose results under the requirement of rapid dose calculation, in this paper, a new interpolation table for dose calculation is proposed utilizing a 3D integral dose calculation method. And the proposed interpolation table was evaluated against the Belgian field experiment. In order to assess the quality of the optimization, the dose results of the new interpolation were compared with that of RIMPUFF, the 3D convolution method and observations from the field experiment. The comparisons indicate that the dose results of the new dose interpolation table are closer to the observed values and have better statistical metrics (0.39 for FB, 0.56 for NMSE, 0.72 for FAC2) than that of RIMPUFF (0.53 for FB, 0.76 for NMSE, 0.65 for FAC2).
The Wear and General Corrosion Properties of Low-Temperature Plasma Nitrided 17-4PH Stainless Steel With a Work-Hardened Sublayer
Authors: Han Zhongli , Lai Ping , Zhuang Wenhua , Guo Xianglong , Zhang Lefu
Published: November 2022

17-4PH stainless steel is used as the material of leadscrew in the reactor control rod drive mechanism (CRDM) in nuclear power plants (NPP). To improve the poor tribological properties without affecting corrosion of 17-4PH stainless steel in-service, the wear and general corrosion properties of low-temperature plasma nitrided 17-4PH stainless steel with a work-hardened sublayer were investigated. Low-temperature plasma nitriding for specimens with work-hardened sublayer was carried out at 390 °C for 1 h, 30 h. The wear test and the general corrosion test were conducted in flowing water at 280 °C and 13 MPa. The general corrosion test was carried out for 2000 h. The results show that the wear volume of the un-nitrided specimen with a work-hardened sublayer varies slightly compared to the unnitrided specimens with a polished surface (without a work-hardened sublayer), but the weight loss of general corrosion is greatly reduced (by a factor of about 3). The wear volumes of the 390 °C / 1 h treated specimen and the 390 °C / 30 h treated specimen are reduced by 3.8 and 4.4 times compared with the non-nitrided specimens (with a work-hardened sublayer), respectively, but the weight loss of general corrosion increases by 1.56 and 1.66 times, respectively.
Study on Selection Method of Shielding Nuclides in Nuclear Reactor Based on Non-Dominated Sorting
Authors: Li Rui , Liu Shichang
Published: November 2022

In the reactor, the main purpose of radial shielding was to protect the internal structure and staff from radiation damage. The good shielding material must have both a high elastic scattering cross-section and a high absorption cross-section to reduce the radiation level outside the reactor core and ensure the safety of workers and equipment outside the reactor. In this paper, a new type selection scheme of shielding material was proposed, which read the reaction cross-sections of all-natural stable nuclides in the ENDF/B-VII database, calculated the single group cross-sections of various reactions of all nuclides by the integral method according to the neutron energy spectrum. Then ranked the nuclides with different cross-sections as the target, to screen the nuclides with the best neutron and photon shielding effect. In this paper, a typical Monte Carlo particle transport program was used for preliminary verification, which was applied to the energy spectrum of the pressurized water reactor and sodium-cooled fast reactor. The results show that for neutron shielding, photon shielding, and neutron-photon shielding, in the thermal neutron spectrum, the coincidence rates between the non-dominated sorting scheme and the Monte Carlo method were 95 %, 98.95 %, and 94.87 % respectively, and in the fast neutron spectrum, the coincidence rates between the non-dominated sorting scheme and the Monte Carlo method were 72.5 %, 95.79 %, and 82.5 % respectively. The calculation time of non-dominated sorting was less than 3% of that of the Monte Carlo method. It provided an efficient method for the selection of shielding nuclides in advanced reactors and had engineering application significance.
Design Overview for Process System of Chlorine Production by Electrolyzing Saline in the Nuclear Power Plant
Authors: Wang Rui , Zeng Kehao , Xu Kechen , Yan Hongmei , Bai Wei , Li Jing
Published: November 2022

A nuclear power plant adopts electrolyzing saline to prepare sodium hypochlorite solution with the effective chlorine concentration no less than 8000 mg/L, which is provided to the circulating water system, essential service water system, industrial water system, potable water system, and demineralized water production system as biocide. The system name of sodium hypochlorite solution production by electrolyzing saline is the circulating water treatment system, which includes salt storage, dilute saline preparation, sodium hypochlorite solution production, sodium hypochlorite solution storage and dosing, etc. In system process design, the system automatic operation is realized by optimizing the system design and operation control scheme, and setting online monitoring instruments, automatic valves and control system to reduce the intervention of operators. In addition, the hydrogen exhaust fans are set next to the outdoor sodium hypochlorite storage tanks for mechanical hydrogen exhaust, so that the hydrogen can be quickly diluted and discharged to the atmosphere to ensure the safety of system operation. Combined with the engineering practice, this paper expounds the design overview of circulating water treatment system from the aspects of user requirements, optimization design of each process link and process equipment configuration, which can provide reference for the process design of circulating water treatment system in subsequent nuclear power plants.
Flow-Induced Vibration of Two Square Cylinders With Rounded Corners in a Tandem Arrangement
Authors: Wu Yidong , Shi Li , Wu Xinxin , Wang Xiaoxin , Xiao Qiankun
Published: November 2022

Based on the design of the heat transfer tube bundles with square cross section of the steam generators in high temperature reactor-pebble bed modules (HTR-PM), a free-vibration experiment is conducted to examine flow-induced vibration (FIV) characteristic of two identical rounded square cylinders with r/D = 0.12 in tandem arrangements at spacing ratio L/D = 1.5∼5.5. One of the cylinders is two-dimensional, spring mounted, and allowed to vibrate in the cross-flow direction while the other is held stationary. Considering two cases that elastic cylinder located upstream or downstream, two mass-damping ratios m*ζ are chosen to investigate the effect of spacing between two cylinder. The vibration responses and the flow structure around two cylinders are studied, using laser displacement sensor and Particle image velocimetry. It is observed that only vortex-induced vibration (VIV) occurs when the elastic cylinder is located downstream. When elastic cylinder is located upstream, the full interaction between VIV and galloping of cylinder with lower m*ζ is invariable no matter what the spacing is. The spacing has noticeable effects on the vibration behavior of cylinder with higher m*ζ and changes the interaction between VIV and galloping. The flow structure of two cylinders indicates that the upstream shear layer reattaches on the surface of the downstream cylinder as L/D < 3.5, where the St drops with increasing L/D. There is gap vortices between two cylinders at L/D > 3.5 where the St rises and is close to the value of single cylinder with a larger spacing.
Study on the Growth and Scintillation Properties of CLLB Crystals
Authors: Wu Kun , Zhang Leilei , Li Haijun , Huang Guangwei , Li Jiaming , Wang Libin , Zhang Siyuan , Li Zhiyuan , Wei Qinhua , Zhou Chunzhi , Wang Zungang , Liu Huilan , Zhu Hongying , Song Yushou
Published: November 2022

Cs2LiLaBr6:Ce (CLLB) scintillation detector has excellent energy resolution and high light output to gamma-rays, with excellent neutron/gamma-rays discrimination performance. Its applications are mainly limited to crystal growth with large size and high quality. In this paper, high performance CLLB was grown, encapsulated and tested. The vertical Bridgman method was employed to grow Cs2LiLaBr6:1%Ce crystals. By regulating the temperature, the maximum temperature gradient of the growth furnace was adjusted to the position of crystal precipitation. To avoid the non-uniform melting component area, the non-stoichiometric of raw materials increasing the excess of 100% LiBr were used. The crystals was developed successfully with 1 inch diameter, 5 cm length, and 4.32 g/mL density. Finally, the crystal was encapsulated and tested, with an energy resolution of 4.0% and relative (NaI) optical output of 1.14 to 0.662 MeV gamma-rays, and an absolute optical output of 47665 Photons/MeV to 0.511 MeV gamma-rays; The figure of merit (FOM) of the neutron/gamma-rays discrimination is 1.27 enabled by charge comparison method. The scintillation decay time of the two kinds of signals is measured respectively by the optical flux simulation method, with a fast component of about 0.19 μs and a slow component of 1.08 μs. Results show that the neutron signals has a higher fast component proportion to gamma-rays signals.
CFD Analysis of the Steam Side in a Cross-Flow X-Type Steam Superheater
Authors: Zhao Shujie , Li Weihua , Hao Wentao , Zhang Wenwen
Published: November 2022

A three-dimensional (3D) computational fluid dynamics (CFD) investigation of steam flow distribution of a cross-flow X-type steam superheater in a small modular reactor (SMR) has been carried out, and different flow distributors have been compared. As a non-baffle shell and tube heat exchanger, the secondary steam flow of the superheater flows across the tube bundles and is superheated in the shell side. The flow distributor is arranged between the shell side inlet and the tube bundles, devices compared are the impingement plate and orifice plate. The diameter and the geometry distribution of the holes on the orifice plate have also been researched. The software used is ANSYS FLUENT, applying standard k-epsilon turbulence flow model. Distribution of velocity and pressure drop of the shell-side steam is obtained. As the analysis of the flow-field distribution contributes to optimizing the heat exchange performance, the analysis solutions can provide helpful information for the superheater design.
Development and Preliminary Verification of Multi-Group Cross Section Processing Code MAGIC for Fast Reactor Based on Continuous Point Cross Section
Authors: Zhang Bin , Ma Xubo , Hu Kui , Zhang Teng , Huang Yuqin , Ma Xuan
Published: November 2022

To meet the needs of fast reactor high precision nuclear data processing, a high precision multi-group cross section processing program MAGIC based on continuous point cross section is developed, which can provide accurate multi-group cross section for fast reactor deterministic program. The MAGIC program uses 1/120 lethargy width to divide the energy interval, and a total of 2082 energy groups are divided in 0.414eV∼14.2MeV for fast reactor energy region. The main functions of the program are as follows: (1) by using the problem dependent neutron spectrum, the effect of resonance interference can be well considered. (2) The resonance self-shielding factor iteration method can be used to better consider the interaction of the resonance self-shielding in the resolved and unresolved resonance energy regions of multiple nuclides. (3) When calculating the elastic scattering matrix, the resonance phenomenon is considered in the one-dimensional elastic scattering cross section, and the prefabricated scattering transfer probability table can greatly improve the calculation efficiency of the scattering matrix under the ultra-fine group energy framework on the premise of maintaining the accuracy. In terms of benchmark verification, the effective increment coefficient was verified by fast energy spectrum critical benchmark based on the multi-group cross section generated by MAGIC. The difference between the calculated results and those of the Monte Carlo program and NJOY was small, which preliminatively verified the effectiveness of the multi-group cross section generated by MAGIC program in the calculation of deterministic method.
Eulerian Grid Approximation Algorithm of Droplet Motion in Gravity Field
Authors: Liu Zhanwei , Bo Hanliang
Published: November 2022

Accurate and efficient gas-liquid two-phase flow simulation is of great significance to energy, chemical industry, aerospace, environment and other fields. Although the Euler-Lagrange method can accurately simulate the motion of droplets, its computational complexity increases significantly with the number of droplets, thus it’s difficult to be applied to practical applications. Therefore, based on the explanation of the physical mechanism of the droplet movement in the steam-water separation device, combined with the Euler-Lagrange model, the Euler grid approximation model is established, which can simultaneously calculate the information of the whole droplet field and reduce the amount of computation while maintaining accuracy. In this model, the Lagrange description of droplet motion is approximated by Euler grid approximation method, and the real droplet motion is approximated as the motions of droplets along the grid nodes, so that the line between nodes is used to simulate the flow line. This paper gives the physical mechanism explanation and mathematical expression of the Euler grid approximation model to obtain the velocity, density, rotation under the gravity field, and the accuracy of the method increases with the refinement of the grid. The Euler grid approximation method can be applied to droplet motion subject to body force, as well as multi-droplet motion.
Preliminary Development of a Coupling Environment Based on MOOSE and OpenFOAM and Its Application on Plate Fuel Modeling
Authors: Liu Hongquan , He Yanan , Wu Yingwei , Su Guanghui , Tian Wenxi , Qiu Suizheng
Published: November 2022

High-fidelity analysis of nuclear reactors is beneficial to improving the efficiency of nuclear reactor designs. To enhance modeling fidelity, multi-physics coupling simulations are typically employed. To this end, several platforms for code coupling are under active development, including MOOSE, SALOME, et al. In this paper, the loosely coupling interface between MOOSE and OpenFOAM was developed, aiming to extend the Computational Fluid Dynamics (CFD) capacity for MOOSE. Specifically, process control, data transfer, and mesh projection of the coupled simulation are realized by the “MultiApps” and “Transfers” module within MOOSE and the “externalCoupled” module within OpenFOAM, where text-based data transfer is used. The CFD capabilities of OpenFOAM and the finite element analysis capabilities of MOOSE are fully utilized in this coupling environment. Subsequently, a conjugated heat transfer problem for plate fuel was conducted to demonstrate the feasibility of the developed interface, where variables consisting of temperature, convective heat transfer coefficient, and heat flux at outer surfaces are transferred between MOOSE and OpenFOAM. The predicted results of the plate fuel seem reasonable and the validity of the novel coupling code for MOOSE and OpenFOAM is preliminary confirmed.
Study on Radioactive Inventory, Accident Source Term and Offsite Dose Consequence of DHR400 Under Fuel Handling Accident and Core Damage Accident
Authors: Wang Fenglong , Mei Xuanlan , Zhang Liang , Yang Yong , Zhang Qiang , Wang Shixi , Wu Mingyu
Published: November 2022

The 400MWt district-heating reactor (DHR400), which is one type of small modular reactor (SMR), is designed to solve district-heating problem in northern China, and its reactor characteristics and parameters are described in the paper. Reactor core inventory, accident source term and dose consequence of DHR400 is calculated in the event of fuel handling accident and core damage accident. The ORIGEN2 code developed by Oak Ridge National Laboratory is used to calculate the core inventory for an operating period of five years before refueling. Based on regulations for environmental radiation protection of China and regulatory guide of NRC, the accident scenario and source term assumption of design based accident and beyond design based accident are discussed. The accidental releases of radioactive material from reactor is calculated and evaluated based on the ARCON96 program sponsored by the NRC. In this paper, an overall systematic analysis method, which includes key aspects in the calculation, such as hazard assessment, source term estimation and timing, dose assessment and so on, is proposed to determine emergency planning zone (EPZ) of DHR400. The result shows that 114m could be used as the recommended boundary of EPZ of DHR400 in one specific DHR400 site.
Experimental Study on Statistical Properties of Two-Phase Flow Instability for Boiling
Authors: Sun Yuanli , Song Zhihao , Song Houde , Liu Xiaojing , He Hui
Published: November 2022

The study of two-phase flow instability is of interest in all fields encountering flow boiling. Compact systems relying on flow boiling due to its superior transport capabilities of heat and mass are prone to several modes of flow instability, which adversely affect the safety and performance of systems. As such, understanding of the flow instability is critical to the design and operation of compact flow boiling systems. In current paper, the instability caused by channel-to-channel interaction is investigated experimentally in a test section with 6 parallel small rectangular channels having hydraulic diameter of Dh = 526.2 μm by using deionized water as working fluid. Boiling leads to very asymmetrical flow distribution within the 6 channels, which results in the simultaneous existence of different flow regimes along the transverse direction of test section. Time traces of pressure drop for different channels and operation conditions are analyzed to explore the properties of channel-to-channel interaction by using spectral analysis. Bubble dynamics including nucleation and growth give rise to flow fluctuations in individual channel, and the time intervals between flow fluctuations are caused by a series of time-scale self-convolutions of bubble duration. Based on the aforementioned mechanism, the time intervals between flow fluctuations can be predicted statistically by the gamma distributions.
Application and Prospect of Additive Manufacturing Technology in Nuclear Fuel Assembly Fabrication
Authors: Ding Jie , Zheng Yixiong , Ye Chen , Li Cong
Published: November 2022

With the advent of additive manufacturing technology and the rapid development of technology in recent years, the design ideas and manufacturing methods of special-shaped and complex components have been greatly broadened. Combined with the continuous improvement of digital design and manufacturing tools, additive manufacturing technology has been widely used in aviation, aerospace, automotive and other industrial fields. Due to the high safety and reliability requirements for nuclear power equipments, the application of additive manufacturing technology in this field is still in the trial application stage before large-scale engineering promotion. This paper focuses on nuclear fuel, which is the core equipment of pressurized water reactor nuclear power plants. Based on the analysis of current domestic and foreign application examples, the key technologies of additive manufacturing of nuclear fuel are preliminarily explored, and the problems and challenges in the future engineering applications are prospected.
Research on Heat Transfer Performance of Heat Pipe in Mobile Small Reactor
Authors: Lu Huaichang , Zhou Tao , Liu Wenbin , Tang Jianyu , Gao Tianyu
Published: November 2022

Mobile heat pipe small reactors, with their simple structure, high mobility and inherent safety features, is a hot frontier topic for research at present. Based on Fluent software, numerical simulations are used to study the changes in the steady-state operating parameters of a high-temperature potassium heat pipe, and to investigate the effects of the horizontal acceleration, inclination angle and saturation temperature of the mass inside the heat pipe on the overall heat transfer performance of the heat pipe. It is found that the heat transfer performance of the heat pipe increases as the saturation temperature of the mass inside the tube decreases; the overall thermal resistance of the heat pipe tends to decrease and then increase slightly with increasing horizontal acceleration; the overall thermal resistance of the heat pipe tends to decrease and then increase with increasing tilt angle.
Fluid-Thermal-Mechanical Coupling Analysis of the Reactor Vessel of Natural Circulation Lead-Cooled Fast Reactor SNCLFR-100
Authors: Duan Wenshun , Zhang Kefan , Chen Hongli
Published: November 2022

Lead-based reactors mostly adopt integrated pool structures to simplify the design of reactor and have high inherent safety. As an important safety facility in the primary circuit, the reliability of the reactor vessel is critical for the safe operation of the whole reactor. The reactor vessel of natural circulation lead-cooled fast reactor SNCLFR-100 was taken as the object, and the corresponding simplified model was established. Based on ANSYS, the fluid-thermal-mechanical coupling analysis was carried out under the steady-state to study the influence of flow field and temperature field on the reactor vessel, and to reveal the distribution law of key thermal and mechanical parameters under the steady-state operation in this paper. The results of mechanical calculations were evaluated based on American society of Mechanical Engineers (ASME) standards, and the preliminary design scheme of reactor solid structure was optimized. The results showed that the stress evaluation results of the SNCLFR-100 reactor vessel meet the requirements of ASME standards in steady-state operation conditions. After optimizing the original solid structure, the stress evaluation results show that the maximum stress value was reduced by 43.3%, and the stress concentration was obviously suppressed, which indicates that the reactor container has a higher safety margin under the optimization scheme. The Steady-state calculation based on normal operation conditions provides a certain research basis for the subsequent transient calculation under accident conditions.
Design and Validation on Preload of Manway Bolt for Pressurizer of Nuclear Power Plant Under Hydrotest Condition
Authors: Chen Tao , Xiong Guangming , Fu Guanhua , Zhao Xiaolong , Wang Defu , Zhang Yongqiang , Liang Enming
Published: November 2022

Bolted-flange connected structure is widely used on Primary Equipment of Nuclear Power Plant and it is one of the most important assemblies of Primary Coolant Loop. Special attentions should be paid for the design and validation of this kind of structure for high risks of leakage and structure integrity of the equipment. In this paper, preload of manway bolt for Pressurizer of Nuclear Power Plant under hydrotest condition was calculated theoretically. Bolt-load distribution for the bolted-flange connected structure was simulated by Finite Element Analysis method with the above preload applied. Bolt load loss and asymmetric distribution during the test was presented resulting from the simulation for saddle-shaped manway flange. Assessment was carried out according to ASME in the aspect of stress in the bolt shank, pressure on the gasket and rotating angle of the flange (sealing verification) with the corrected preload applied. The simulation results were validated by 1:1 prototype test carrying out for the Pressurizer and they were consistent with the test results. The value of coefficient Ks for bolt preload design formula was confirmed by the test. Key aspects of influence on manway bolt preload design were summarized and full design formula as well as value of the design preload for hydrotest condition were given.