核动力工程

Key Technology of ACP100: Reactor Core and Safety Design

Song Danrong, Li Qing, Qin Dong, Dang Gaojian, Zeng Chang, Li Song, Xiao Renjie, Wei Xuedong

2021, 42(4): 1-5. doi: 10.13832/j.jnpe.2021.04.0001

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Abstract:
Small modular reactor is a new kind of nuclear energy system. The ACP100 is a multi-purpose modular small PWR with full intellectual property in China. This paper introduces the research and development process, the main characteristics of the reactor core and safety design technology, mainly including the nuclear design, thermal-hydraulic design, safety design concept, inherent safety design, and the strategy for accidents. Through the combination of the deterministic theory and the probabilistic safety assessment, the safety of ACP100 is greatly improved and exceeds the Generation 3 nuclear power plant safety standards.

Research on Characteristics of U-tube Backflow in Natural Circulation Steam Generator under Ocean Conditions

Li Xiaojia, Zhang Yong, Cong Tenglong, Li Peiying, Lu Chuan, Zhang Jibin

2021, 42(4): 6-13. doi: 10.13832/j.jnpe.2021.04.0006

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Abstract:
When the pressurized water reactor is in natural circulation, reverse flow may occur in the U-tube of the steam generator, which will increase the flow resistance of the primary circuit and decrease the flow of natural circulation, which will adversely affect the safe operation of the reactor. Based on the RELAP5 program, the additional force model under ocean conditions and the spatial coordinate solution model of the control body are established, all U-tubes of the steam generator are modeled and node division are conducted, and the steam generator under ocean conditions (inclination, undulation, swing) is calculated. The critical mass flow rate of the reverse flow and the pressure difference between the inlet and outlet of the inner U-shaped pipe are analyzed. Finally, the influence of three ocean conditions on the reverse flow of the fluid in the U-shaped pipe is analyzed. The results show that the backflow phenomenon may be changed under inclined conditions; the undulating conditions that may be encountered during navigation cannot change the backflow phenomenon; the backflow phenomenon may be changed when the sway conditions are severe.

Molecular Dynamics Simulation on Microscopic Characteristics of Carbon Dioxide in Trans-Critical Progress

Tang Jia, Huang Yanping, Wang Junfeng, Zang Jinguang, Liu Guangxu, Liu Ruilong

2021, 42(4): 14-20. doi: 10.13832/j.jnpe.2021.04.0014

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Abstract:
Molecular dynamics (MD) simulations was performed to analyze the microscopic characteristics of carbon dioxide in trans-critical progress. Radial distribution function (RDF) analysis exhibited that the short-range structure varied weakly near the critical point, which was mainly enhanced by the strong neighbor intermolecular interaction. The simulation results of the coordination number in first shell further show that the variation of short-range structure primarily lies in the number change of coordinated molecular; The gaseous CO₂ is still of an ordered structure in the short-range and of a disordered structure in the long-range; The static structure factor analysis showed the existence of medium/long-rang ordered structure. Disorder range was defined and its abrupt surge showed explicitly that the range of intermolecular interaction increased in the pseudo-critical region.

Numerical Investigation on Thermal Hydraulics of Helical Coil Tube Once Through Steam Generator for LBE Fast Reactor

Ding Xueyou, Chen Zhiqiang, Wen Qinglong, Ruan Shenhui, Qiao Pengrui

2021, 42(4): 21-26. doi: 10.13832/j.jnpe.2021.04.0021

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Abstract:
The design of Lead and Bismuth Eutectic (LBE) cooled fast reactor helical coil tube steam generator is modeled in this study. Physical modeling method is developed with the combination of fine grid blocks and porous medium. Through the coupling the primary three-dimensional turbulence calculation and the secondary side heat transfer based on UDF (User Defined Function) method, a numerical analysis of the thermal hydraulic characteristics of the steam generator is carried out in the FLUENT solver. The results show that: a. the flow distribution holes near LBE inlet and the straight tubes in the chamber have peaks in the flow rate of LBE, and the maximum radial velocity is 0.431 m/s, b. when LBE flows from the inlet chamber into the tube bundle, the pressure, cross flow and axial flow rates change rapidly because of resistance mutation, and c. the changing process of thermal hydraulic parameters confirms to the qualitative mechanism analysis results of flow and heat transfer. The maximum temperature difference between primary and secondary sides near the Critical Heat Flux (CHF) point is 109.61 K, and the maximum heat flux is 323.55 kW/m². This study will provide an important reference for HOTSG structure design, fluid-induced vibration and safety evaluation of lead-bismuth fast reactor.

Evaluation of Condensation Heat Transfer Coefficient of Stable Steam Jet Submerged in Water

Wang Jue, Chen Lisheng, Liu Jiange, Hu Chen, Cai Qi

2021, 42(4): 27-32. doi: 10.13832/j.jnpe.2021.04.0027

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Abstract:
In order to analyze the heat transfer characteristics of a stable submerged steam jet, three types of condensation heat transfer coefficients (HTCs) were evaluated. The results show that the accuracy of the experimental values of the average HTC is mainly influenced by the calculation of the interfacial area, and the traditional semiempirical correlation (characterized by the condensation driving potential and the steam mass flux) has a large deviation from the prediction at different discharge diameters. A fully-empirical correlation with a wider applicable range can be obtained by adding the discharge diameter as independent fitting variable, and the discrepancy between prediction and experimental data is within ±30%. The accuracy of the interfacial HTC is mainly influenced by the microscopic parameters of the steam plume. The dimensionless HTC (characterized by the dominant frequency of pressure oscillation) deviates significantly from the experimental value at low water subcooling, and the predicted trend is similar to the experimental value when the steam plume penetration length is fitted in the correlation.

Experiment Study of Heat Transfer to Supercritical Water in a Triangular-Lattice Configuration

Cui Dawei, Chen Shuo, Gu Hanyang, Lin Jiming

2021, 42(4): 33-38. doi: 10.13832/j.jnpe.2021.04.0033

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Abstract:
The experimental research of heat transfer to supercritical water in a triangular-lattice configuration has been performed on the supercritical water multipurpose test loop. Circumferential non-uniformed wall temperature distribution and heat transfer enhancement induced by the grids were observed in the bundles. The heat transfer data under conditions of different heat flux, mass flux and pressure were obtained. Finally, an empirical correlation with the prediction deviation of ±15% was developed to predict the supercritical heat transfer behaviors in the triangular-lattice.

Steady-State Thermal-Hydraulic Analysis of Steam Generator with Axial Economizer Based on Node Method

Su Shu, Liu Chengmin, Huang Wei

2021, 42(4): 39-44. doi: 10.13832/j.jnpe.2021.04.0039

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Abstract:
Based on the one-dimensional flow hypothesis, heat transfer hypothesis, and two-phase thermal equilibrium hypothesis, a lumped parameter method and a distributed parameter method are used to establish a one-dimensional steady-state thermal-hydraulic analysis model of the steam generator with axial economizer. Using C++ language programming, the calculation results are compared with the design values of the thermal-hydraulic parameters of a typical steam generator with axial economizer. The results show that the relative error of most of the overall parameter calculation results is within 3%, which verifies the rationality of the model. The analysis of the temperature, cavitation share, pressure and other parameters of the steam generator along the direction of the measured fluid flow trend consist well with the results of thermal hydraulics and qualitative mechanism analysis, indicating that the model and solution method established in this paper can accurately predict the steady-state thermal hydraulic parameter distribution of the steam generator with axial economizer.

Development of Device for Measuring the Heat Generation Rate of Materials in Research Reactor

Zhao Wenbin, Yang Wenhua, Nie Liangbing, Si Junping, Xu Bin, Sun Sheng, Tong Mingyan

2021, 42(4): 45-50. doi: 10.13832/j.jnpe.2021.04.0045

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Abstract:
To explore the axial distribution of the heat generation rate of the materials in the channels of the test reactor, a heat generation rate measuring device was designed, by taking the G7 channel of the high flux engineering test reactor (HFETR) as an example. The strain distribution contours of the device and test section under load are obtained by numerical simulation method, and the temperatures of the calorimeter to calibrate the bridge and the measurement bridge are obtained by physical calculation. The heat generation rate measurement is carried out in the G7 channel using this test device. The results show that the overall structure of the device meets the strength requirements, and a protective tube needs to be installed between the calorimeters in the test section. The calculated temperature of the sample and the bridge is lower than the melting point of the material, and the device meets the thermal requirements. The heat generation rate measured by the test varies with the reactor power changes regularly. Different materials have different γ absorption properties under different energy levels of γ-ray environment. The device can be used as a heat generation rate measurement tool to provide a guarantee for determining the heat generation rate distribution of different materials in the nuclear reactor.

Study of Boiling Heat Transfer Model of Confined Bubble Flow and Annular Flow in a Heating Rectangular Mini-Channel

Yu Zhongbin, Li Yi, Tian Ye

2021, 42(4): 51-55. doi: 10.13832/j.jnpe.2021.04.0051

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Abstract:
the Based on the typical characteristics of the restricted bubble, the geometric structure of confined bubble flow is investigated. Based on the time weighted average method, the weights of the confined bubble and the liquid plug area in the confined bubble flow are determined. Based on one-dimensional heat conduction theory and integral method, a method for calculating the evaporation heat transfer coefficient of liquid film was established and applied to the annular flow area.The model works with Re range of 2300 to 5373, Pr range of 2.75 to 19.8, and Ca range of 0.000835 to 0.002767.

Effect of Nucleation Density Model on CHF of Curved Surface

Li Dan, Yang Daibo, Li Kun, Li Gang, Jia Yige, Yao Zhang, Li Ang

2021, 42(4): 56-62. doi: 10.13832/j.jnpe.2021.04.0056

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Abstract:
In the event of a serious accident, external cooling can be applied to the lower head of the reactor pressure vessel (RPV) in order to reduce the possibility of damage to the lower head. However, there will be great heat flow surrounding the lower head of the RPV, so external cooling may cause subcooled boiling, which gathers bubbles and deteriorates heat exchange, even burns out. This research uses ANSYS Fluent to calculate the critical heat flux (CHF) for external cooling of the RPV, and it is found that the nucleation density model studied by Basu Warrier and Dhir can be well applied to the calculation of CHF on the spherical surface. By comparing the CHF of the spherical and ellipsoidal lower head, it is believed that the CHF characteristics of the ellipsoidal lower head are completely different from the spherical structure. The experimental and calculation results of the spherical structure cannot be used to infer the numerical value and variation of the ellipsoidal structure.

Study on Calculation Method of Courant Limit in Thermal Hydraulic System Analysis Code

Li Jiangkuan, Huang Tao, Lin Meng, Wang Xu, Chen Junjie

2021, 42(4): 63-67. doi: 10.13832/j.jnpe.2021.04.0063

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Abstract:
In order to analyze the influence of different Courant limit calculation methods on the calculation speed and accuracy of reactor thermal hydraulic system analysis code, two kinds of Courant limit calculation methods were studied: synthesis method and grouping method. The calculation principles of the two methods were analyzed, and the two methods were adopted to calculate the steady-state condition and large break loss of coolant accident condition of a PWR respectively. The results show that there is no significant difference between the two methods in steady-state condition; in the case of large break loss of coolant accident with drastic change of velocity field, the synthesis method can obtain more accurate calculation results, but it takes more time, and the grouping method can obtain faster calculation speed, however its calculation accuracy is lower.

Analysis and Solution to Abnormity of DNBR Indicated by ICIS of a WWER Unit

Fang Jun, Yang Changjiang

2021, 42(4): 68-72. doi: 10.13832/j.jnpe.2021.04.0068

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Abstract:
To analyze the reason of difference between the departure from nucleate boling ratio (DNBR) indicated by the upper-level software and lower-level software of the in-core instrumentation system (ICIS) at a certain water-cooled water-moderated power reactor (WWER) unit under its first thermal test during raising power, an investigation was made on the critical heat flux (CHF) correlations adopted in the reactor thermal and hydraulic design and accident analysis of WWER unit. On the above basis, the reason behind the difference was found to be the use of different CHF correlations by the upper-level software and the lower-level software of ICIS through the simulation of thermal tests under 50%, 75% and 90% power level using WWER accident analysis code DINAMIKA-97, calculation of DNBRs and comparison with the ones measured in thermal-hydraulic tests. DNBR under 100% power level was predicted, which coincided with the measured DNBR very well and proved the correctness of the guess further. It is suggested that the CHF correlations adopted by the upper-level software and the lower-level software shall be modified to the same conservative CHF correlation to get a conservative DNBR.

Study on Molecular Dynamics of Singular Nature of Physical Properties near Critical Point in Carbon Dioxide System

Tang Jia, Huang Yanping, Wang Junfeng, Zang Jinguang, Liu Guangxu, Liu Ruilong

2021, 42(4): 73-79. doi: 10.13832/j.jnpe.2021.04.0073

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Abstract:
Based on the molecular dynamics simulations, this paper analyzed the singular nature of physical properties of the carbon dioxide near the critical point. Detailed analysis of density simulation exhibited that an excellent agreement can be obtained by COMPASS forcefield besides the pseudo-critical region, while the peculiar behavior of density can again be observed. The computational domain is divided into the small sub-domains in which the standard deviation are evaluated as density fluctuation, and it shows maximum near the critical point and higher values at supercritical condition than that at subcritical condition. A new 3-parameter model was put forward to define the CO₂ dimer distribution after each molecule in the domain had been supplied with the local coordination frame. The result shows that the T-shaped dimer is with higher probability to be formed than the parallel and crossed configuration, and the conversion of these dimer conformations closely associates with the peculiar properties of the critical point.

Study on Gravity Sedimentation of Multicomponent Hygroscopic Aerosols in Reactor Severe Accident

Lu Junjing, Mao Yawei, Zhang Tianqi, Zhu Bolin, Yang Xiaoming, Ma Rubing

2021, 42(4): 80-85. doi: 10.13832/j.jnpe.2021.04.0080

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Abstract:
In the event of a severe accident, the multicomponent hygroscopic aerosols in the containment will absorb water under the high humidity condition, thereby influencing the gravity sedimentation behavior. In this study, a physical model of the equilibrium particle diameter of the multicomponent hygroscopic aerosol particles was developed through theoretical analysis, and it was also validated by experimental results. The model focuses on the effect of solubility on the hygroscopic process and explains the reason why the multicomponent hygroscopic particles grow along a discontinuity curve. Based on a typical gigawatt-class pressurized water reactor, the effects of relative humidity, dry particle diameter and mass fraction of hygroscopic components on the removal coefficient of the gravity sedimentation were investigated. The results show that the velocity of the gravity sedimentation will significantly increase, only if the aerosol particles grow to a certain degree. Only when the humidity is more than a certain value, the sedimentation process of the pure hygroscopic aerosol particles with a dry particle diameter exceeding 0.1 μm will accelerate due to hygroscopicity, and this humidity limit will decrease as the dry particle diameter increases. With the progress of the accident, the mass fraction of non-hygroscopic components in the aerosol particles gradually decreases, leading the above-mentioned humidity limit increasing and the acceleration of gravity sedimentation due to the hygroscopicity decreasing in the same humidity.

Experimental Investigation on Plume Length of Submerged Steam Jet through Spargers

Wang Jue, Chen Lisheng, Liu Le, Hu Chen, Zhang Wei

2021, 42(4): 86-90. doi: 10.13832/j.jnpe.2021.04.0086

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Abstract:
The direct contact condensation of saturated steam with a mass flux between 300 and1100 kg·m⁻²·s⁻¹ submerged in the subcooled water at a temperature of 35 to 65℃ was studied experimentally with I-type side-opening spargers, which apertures were 4, 10 and 16 mm, respectively. The results show that: when the aperture is fixed, the penetration length of the steam plume increases with the increasing of the steam mass flux and the pool water temperature. The penetration length of the steam plume through a large-opening sparger is close to that through a straight nozzle, and the deviation between the fitting value and the experimental value is within ±15%. The penetration length of the steam plume through a small-opening sparger is obviously lower than that through a straight nozzle, and the deviation between the fitting value and the experimental value is up to 80%. The steam mass flux is re-calculated utilizing the correlation specified for a contraction nozzle to consider the injecting characteristic of an I-type sparger, and the deviation between the fitted value and the experimental value is within ±20%. A new semi-empirical correlation is fitted and the discrepancy between the prediction and the experimental value is within ±10%.

Research on Judgment of Supercritical Water Heat Transfer Deterioration Based on Machine Learning

Ma Dongliang, Zhou Tao, Huang Yanping

2021, 42(4): 91-95. doi: 10.13832/j.jnpe.2021.04.0091

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Abstract:
In order to further improve the safety and stability of supercritical water reactors, avoid the occurrence of the heat transfer deterioration in supercritical water, based on the existing experimental data of supercritical water heat transfer, using several main machine learning algorithms, the classification and judgment and prediction accuracy analysis of the experimental parameter state points of supercritical water were made to determine the occurrence of the heat transfer deterioration. The research results show that the random forest algorithm has the highest average prediction accuracy for the test data, reaching about 97.8%. The average prediction accuracy of the K-nearest neighbor algorithm is the lowest, but it also reaches about 91%. At the same time, the importance of various influence parameters on the selection of heat transfer deterioration was analyzed.The most important parameter related to the heat transfer deterioration judgment is the specific enthalpy, and the second important parameter is the heat transfer coefficient. The third important parameter with contribution to the heat transfer deterioration is the pipe diameter.

An Effective Thermal Model of Coated Particle Dispersed Fuel with High Packing Fraction

Li Wenjie, Yu Hongxing, Xiao Zhong, Jiao Yongjun, Chen Ping, Li Yuanming

2021, 42(4): 96-100. doi: 10.13832/j.jnpe.2021.04.0096

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Abstract:
Accurate prediction of the temperature distribution in nuclear fuel is an important step for the design and screening of multi-coating particle dispersed nuclear fuel. This study aims to establish an effective thermal model and numerical solutions for macro scale heat transfer analysis by investigating the effective thermal properties of multi-coating particles and the dispersed fuel bulk. The effects of particle arrangement, sizes, clustering on temperature distribution inside nuclear fuel element was studied with the established effective model. This study helps clarify the micro- and macro- scale heat transfer mechanism of multi-coating particle dispersed fuel, and provides a guidance for the design, optimization and safety analysis of this type of fuel.

Calculation of Internal Stress in Oxide Films of Zirconium Alloy

Zhang Junsong, Lyu Junnan, Long Chongsheng, Liao Jingjing

2021, 42(4): 101-104. doi: 10.13832/j.jnpe.2021.04.0101

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Abstract:
The internal stress in the oxide film of zirconium alloy is an important factor in the corrosion kinetics of zirconium alloy. At present, there is no unified method to obtain the internal stress in the oxide film, and the data is quite different. Based on the traditional experimental and theoretical methods, the double layer oxidation bending model of ZrO₂/Zr is established, the internal stress in the oxide film under different corrosion conditions is calculated, the change rule of the internal stress is obtained and the influencing factors is analyzed, which provides a more reliable way for the research of the internal stress in the oxide film of zirconium alloy.

Simplified Modelling Method for Spacer Grid of NHR200-II Fuel Assembly

Wang Xicheng, Wang Dingqu, Jiang Yueyuan, Li Songyang

2021, 42(4): 105-111. doi: 10.13832/j.jnpe.2021.04.0105

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Abstract:
A simplified modelling method for the spacer grid of NHR200-II fuel assembly is proposed to increase the computational efficiency in the dynamic analysis. It includes the replacement of the grid inside the dimples and three-arc springs with the non-linear connectors, and the replacement of the fuel rods with beam elements. The stiffness of connectors was derived from previous experiments of NHR200-II spacer grid. Natural frequency analysis and impact simulation were performed by a 1×2 sub-model of the fuel assembly to verify this method. Then, this simplified method was applied to a full-scale 9×9 spacer grid model, to study the effects of the grid clamping ability on its dynamic characteristics. The results suggest that this simplified approach can give a reasonable response when induced by a seismic load at different clamping levels. In conclusion, from the modelling point of view, the proposed simplified method for NHR200-II spacer grid is effective.

Research on Analysis Method for Performance of Fuel Element Based on Thermal-Fluid-Solid Coupling

Huang Yongzhong, Li Quan, Li Yuanming, Pang Hua, Lu Huaiyu, Liu Zhenhai, Qi Feipeng

2021, 42(4): 112-118. doi: 10.13832/j.jnpe.2021.04.0112

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Abstract:
The existed fuel performance analysis tools are not applicable to the hollow prism fuel with special structure and operation conditions, so a new method is needed to assist the fuel design and evaluation. In this paper, a 3D fluid-thermal-solid coupling analysis method was established based on the COMSOL software by conjugate heat transfer technology and the equivalent material property models for particle reinforced composites, and had been verified with the General Electric data. Temperature and thermal stress of fuel elements in different sizes and axial power distributions were calculated with this method. The results show that maximum temperature exists at the side edge of prism, and maximum thermal stress exists at the thinnest inner wall. The thinner and longer fuel has the smaller maximum thermal stress and temperature. Flatting the axial power distribution in the entrance region can decrease the maximum thermal stress and temperature slightly. This analysis method can be used to optimize the design of the hollow prism fuel element.

Study on Corrosion Resistance of ODS-FeCrAl Tube for ATF

Li Jing, Wu Sajian, Yang Ying, Xiong Liangyin, Ma Haibin, Liu Shi

2021, 42(4): 119-125. doi: 10.13832/j.jnpe.2021.04.0119

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Abstract:
Oxide dispersion strengthened (ODS) FeCrAl alloy is one of the candidate materials for Accident Tolerant Fuel (ATF) cladding tubedue to its excellent mechanical strength, creep resistance and radiation-swelling resistance at elevated temperatures. The corrosion behaviors of ODS-FeCrAl tube with 9.5 mm outside diameter and 0.3 mm wall thickness in static water at 360 ℃ and18.6 MPa for 100 days, in flowing aqueous solution containing 1200 ppm B and 2.2 ppm Li at 360℃ and 18.6 MPa for 100 days and in steam at 1200℃ and 0.1 MPa for 8 hours, were studied herein. The morphology, composition and element distribution of oxide film were analyzed by SEM, XPS and XRD respectively. The corrosion products are mainly Fe₃O₄, due to the low oxygen content in both aqueous environments at 360℃. And the weight gain is 0.036 mg/cm² and 0.36 mg/cm², which corresponding oxide film thickness is 0.072% and 0.72% of that of tube wall respectively. In the steam at 1200℃, owing to high temperature and sufficient oxygen content, α-Al₂O₃ oxide film with a mean thickness of 2.34 μm is dominant on the surface, delaying further the oxidation of the matrix.No observable cracks and voids are identified on the surface and the cross section of oxide filmsin all corrosive environments.In comparison with Zr-4 reference cladding, ODS-FeCrAl tube exhibits an outstanding high temperature oxidation and corrosion resistance.

Study of Assembly Nuclide Density Prediction Based on Data Mining Technology

Lei Jichong, Xie Jinsen, Yu Tao, Zhou Jiandong, Chen Zhenping, Zhao Pengcheng, Xie Chao, Ni Zining

2021, 42(4): 126-132. doi: 10.13832/j.jnpe.2021.04.0126

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Abstract:
The DRAGON program was used to calculate 9600 samples, and the nuclide densities of ²³⁵U, ²³⁸U, ²³⁹Pu, ²⁴¹Pu, ¹³⁷Cs, ²⁴⁴Cm and ¹⁵⁴Nd nuclides were used as the prediction parameters. Linear regression model, regression tree model constructed based on decision tree, multilayer perception (MLP) model and random forest model were selected to carry out model training. Pearson correlation coefficient (PCC), mean absolute error (MAE), relative absolute error (RAE) and relative root mean square error (RRSE) were chosen to evaluate the fitting effect of the models; the trained models were used in the test set for the target. The trained models were used to predict the target nuclides in the test set, and their prediction accuracy was evaluated by relative errors. The results show that the training time of the data models is less than 3 s. After the evaluation of the selected parameters, the MLP model has the best training effect among the four models for all the predicted kernels, and its correlation is above 0.999. The average deviation of the MLP model for all the predicted kernels is less than 1%. This paper initially verifies the feasibility of data mining techniques in predicting the density of assembly nuclei.

In-Pile Performance of TRISO Particle Used Stainless Steel Foams as Buffer Layer

Yin Chunyu, Liu Shichao, Jiao Yongjun, Zhou Yi, Gao Shixin, Xing Shuo, Qing Tao, Wang Lida, Yan Xinlong

2021, 42(4): 133-137. doi: 10.13832/j.jnpe.2021.04.0133

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Abstract:
Thermal conductivity of TRISO particle deteriorated during in-pile operation due to the shrinkage of the low density carbon buffer layer and the gap appeared. In order to solve this problem, stainless steel foam was used to replace the low density carbon buffer layer as buffer layer in TRISO particle. The in-pile performance of the new type TRISO particle was simulated, and the result indicated that the stainless steel foam can avoid the gap between the buffer and the IPyC layer which can increase the thermal conductivity and decrease the kernel temperature during operation process; at the same time, regardless of stainless steel foam or low density carbon buffer layer used as buffer layer, the stress on IPyC and OPyC layers all exceeded the strength of the two layers. Hoop stress on SiC layer decreased with the elasticity modulus of metal foam, and the stress can be controlled by decreasing the elasticity modulus. To sum up, metal foam with high porosity and low elasticity modulus can be selected as the buffer layer to improve the thermal conductivity and assure the integrity of the coated layer and increase the life time. This study offered the optimization direction of TRISO particle for engineering application.

Sensitivity Analysis of Multi-Layer Molten Pool Model of PWR Lower Head

Li Zhigang, An Ping, Pan Junjie, Liu Wei, Lu Wei

2021, 42(4): 138-143. doi: 10.13832/j.jnpe.2021.04.0138

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The molten pool model of the lower head is an important model to evaluate the effectiveness of In Vessel Retention (IVR), which has been widely used in the safety evaluation of typical pressurized water reactor (PWR). The distribution and transfer process of the composition and heat of the melt in the pool of the traditional two-layer pool model and the three-layer pool model proposed in recent years are simulated, which is with the characteristics of complex relationship and strong nonlinearity. In order to provide a support for the optimization of the molten pool delamination model and the mitigation strategy of serious accidents, both global sensitivity analysis library (SALib) and IVR analysis code (CISER V2.0) developed by Nuclear Power Institute of China are used to analyze the sensitivity of four molten pool multilayer models and obtain the influence degree of the main input parameters on the key result parameters of each model. The sensitivity analysis results reflected a typical characteristics of the molten pool model. The radius of the lower head has a significant impact on the four molten pool multilayer models, while the input parameters that have significant influence on the key result parameters are basically the same in the Salay & Fichot model and the two-layer molten pool model, the initial mass of molten material has the greatest impact on the Esmaili model, and the density of the molten material has the greatest impact on the Seiler model.

Identification of Correlation among Performance Shaping Factors of SPAR-H Method

Liu Jianqiao, Zhang Li, Zou Yanhua, Sun Qianlin, Liu Xueyang, Chen Shuai

2021, 42(4): 144-150. doi: 10.13832/j.jnpe.2021.04.0144

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Abstract:
Standardized Plant Analysis of Risk-Human reliability analysis (SPAR-H) is an internationally known and accepted human reliability analysis (HRA) method in nuclear power plants (NPPs). However, the eight performance shaping factors (PSFs) overlap, resulting in the double count or over estimation of the human error probabilities (HEPs). To improve its PSFs system, 89 human error event reports related to the operation of operators in the main control room were collected from 219 operating event reports of Chinese NPPs from 2007 to 2017. The correlation among the PSFs was then studied. Therein, three kinds of data mining methods, i.e., association rule analysis, exploratory factor analysis and Pearson correlation analysis, were used. Results show that: a. there is a significant correlation among complexity, stress/stressor, fitness for duty and available time, in which the c the complexity correlates with stress/stressor and fitness for duty, the fitness for duty correlates with stress/stressor, and the stress/stressor correlates with available time; (2) there also exists correlation among work process, procedure, ergonomics/HMI and experience/training. In the event involving the procedure, ergonomics/HMI or experience/training, work process is involved with a high probability. These findings can be used as the reference in the improvement of the PSFs system of SPAR-H and as the basis in the quantitative study of the cause-and–effect dependences among the PSFs.

Demands Analysis on Diagnosis of Nuclear Accident Emergency Status of Chinese Nuclear-Powered Surface Ship

Yu Hong, Cheng Shisi, Li Lan

2021, 42(4): 151-158. doi: 10.13832/j.jnpe.2021.04.0151

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Emergency status classification can reasonably define the emergency response requirements of various emergencies caused by nuclear accidents, to win more time for the adequate but not excessive emergency response. This paper presents the demands on classification matrix and the diagnostic software development with this classification matrix for the diagnosis of nuclear accident emergency status of Chinese nuclear-powered surface ship, through the improvement of a whole set of classification matrix of Unit 1 and Unit 2 in Qinshan No.2 Nuclear Power Plan which represents the overall level of emergency classification technology in China. Firstly, through the establishment of constraints, the classification matrix with strong standardization and logic is developed, especially the ordering of matrixes and their components; and secondly, through the establishment of multi-entry logic tree and user-friendly interface, the diagnosis software with fast diagnosis function is developed, especially the layering of logic trees and the composition of interfaces.

Development and Verification of HLPS Software for LOCA Hydraulics Load Analysis

Li Wenji, Lyu Hong, Zhang Jie

2021, 42(4): 159-165. doi: 10.13832/j.jnpe.2021.04.0159

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In order to analyze the hydrodynamic loading characteristics caused by coolant jetting and pressure relief wave propagating in the reactor primary system of the nuclear power plant during a hypothetical fracture of the reactor primary system, HLPS, a software to calculate the hydraulic load in the loss of coolant accident (LOCA) in PWR primary circuit, is developed on self-reliance using C++ programming language. Taking M310 reactors coolant as the object, the results of HLPS software are compared with engineering data. The results show that the calculated results of HLPS software are in good agreement with engineering data, at the same time, HLPS adopts implicit solution and higher convergence standard, and the result is more accurate. It can be used for the analysis of the primary circuit system coolant loss accident.

Estimation of Inner Wall Temperature in a Two-Dimensional Pipeline for Inverse Heat Conduction Problem Based on Green’s Function

Li Juan, Yin Haifeng

2021, 42(4): 166-170. doi: 10.13832/j.jnpe.2021.04.0166

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Abstract:
It is not allowed to install the temperature sensor through the opening pore in the pipe to measure the inner wall temperature for some pipe systems in nuclear power plants, and thus it is necessary to find an indirect and non-destructive method to obtain the inner wall temperature fluctuations. The inverse heat conduction problem is analyzed based on Green’s function to achieve the inner wall temperature according to the out wall temperature of the pipeline. It is verified by examples and compared with the conjugate gradient method. Results show that the Green’s function method can accurately catch the inner wall temperature fluctuation of the pipe and is applicable for the thicker wall pipe inverse heat conduction problem which is difficult to converge used by the conjugate gradient method. And because there is no need for iteration, the calculation efficient is much higher, which is more suitable for the fatigue monitoring calculation in nuclear power plants.

A Method Study of Numerical Simulation on Flow Field of Spiral Tube in Steam Generator Shell Side

Wang Cong, Zhang Wei, Li Jingsong, Qiao Pengrui, Shi Huilie

2021, 42(4): 171-175. doi: 10.13832/j.jnpe.2021.04.0171

Abstract(315) HTML (214) PDF(52)

Abstract:
The lead based fast reactors often use integrated core design scheme, due to its compact structure, and the helical coil tube type once-through steam generators (OTSG) have been utilized in various reactors. In order to study the flow and heat transfer characteristics of the lead bismuth eutectic based coolant in the shell side of the helical coil tube OTSG in lead based fast reactors, a section-by-section solution method is adopted, and the wall heat flux fitting formula is used to simulate the flow field of shell side, with the help of FLUENT software. Finally, the correctness of the section-by-section solution method is verified, the flow and heat transfer characteristics of the shell side are analyzed, and the calculation data of velocity, temperature and pressure field are obtained. Its results lay a foundation for the next step of steam generator flow induced vibration analysis and high temperature stress calculation.

Root Cause Analysis of In-Situ AP1000 Reactor Coolant Pump Lock Cup Failure

Fan Fuping, Lu Zhiyong

2021, 42(4): 176-181. doi: 10.13832/j.jnpe.2021.04.0176

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Abstract:
The failure of the large-scale canned motor reactor coolant pump (RCP) of Sanmen Unit 2 caused the unit to shut down in the first cycle of operation. In order to analyze the causes of the RCP fault, based on the fault characteristics and cause analysis methodology, this paper develops the RCP fault root cause analysis method. Through the RCP manufacturing record investigation, on-site operation data evaluation, disassembly and inspection forensics analysis, design analysis and test verification, root cause analysis and evaluation, the root cause of the RCP failure is confirmed that the lower thrust runner lock cup is affected by the resonance flow induced vibration. The initial flaw continues to expand under the action of the resonance and eventually separates. The lock cup fragment makes two holes on the stator can and eventually leads to the RCP failure. The root cause analysis method established in this study can be used for the analysis and treatment of similar complex problems.

Development of Underwater Lifting Tool for Spent Fuel Storage Rack in Nuclear Power Plants

Ou Jianlei, Yuan Zhimin, Wu Wei, Luo Wenguang, Liao Jiatao

2021, 42(4): 182-185. doi: 10.13832/j.jnpe.2021.04.0182

Abstract(259) HTML (139) PDF(40)

Abstract:
In response to the requirement for the underwater lifting of spent fuel storage racks in a second-generation nuclear power plant, according to the current situation of the narrow underwater layout of spent fuel storage racks and the analysis of the safety and convenience of underwater lifting interfaces, a grab-and-release rack was developed. Bottom-bottom lattice lifting tool with underwater self-locking lattice is realized by running water holes on the bottom of the rack, and the structure, operation and working principle of this underwater lifting tool are introduced in detail. Finally, the lattice is used in a second-generation nuclear power plant for a lifting operation of 20 rack frames in the factory. The field application shows that the lifting tool is easy to operate and reliable in structure, which meets the requirements for the underwater lifting of the lattice frame.

Reliability Evaluation of Passive System for Nuclear Power Plant Based on Improved Multi-Level Cross Entropy Method

Zhang Yongfa, Jiang Lizhi, Cai Qi, Liu Xiaoya, Jiao Meng

2021, 42(4): 186-190. doi: 10.13832/j.jnpe.2021.04.0186

Abstract(175) HTML (59) PDF(25)

Abstract:
In view of the difficulty in constructing the importance sampling density in the reliability evaluation of passive system and the defects of the standard multi-level cross entropy method in practical engineering application, an improved multi-level cross entropy method is proposed by improving the algorithm structure and introducing the Halton sequence sampling method with better uniformity. Taking an experimental facility for passive residual heat removal system of a certain marine nuclear power plant as an example, the performance verification analysis and example analysis of the IMCE method are implemented. The calculation results show that, the relative error distribution of the improved method is more convergent and the distribution of the coefficient of variation is similar, and the improved method is with better estimation accuracy and robustness with less computation. Moreover, the improved method is with stronger applicability in practical engineering application, because there is no need to set additional smoothing parameter, and the evaluation process can be completed earlier according to the system characteristics and sampling conditions.

Surface Defect Detection on Nuclear Power Plant Components Based on Photometric Stereo under Near-Field LED Light

Huang Sanao, Li Ming, Xu Ke, Shi Yingjie

2021, 42(4): 191-197. doi: 10.13832/j.jnpe.2021.04.0191

Abstract(483) HTML (208) PDF(44)

Abstract:
To improve the visual inspection of nuclear power plant components, this study introduces the technology of three-dimensional reconstruction framework based on photo metric stereo under near-field LED lighting. An iterative algorithm is proposed to estimate the K value of different light sources and the light6 intensities for various image pixels. Combined with the calculation method of the spatial position of the light source and the detected surface points, the light intensity and light direction of different points on the detected surface are estimated under the illumination of near-field LED point light source. A surface defect detection system is designed based on this technology, and it is verified by experiments on surface damage samples and actual nuclear power equipment. The results show that this system can obtain three-dimensional surface data, and for scratch-type defects, it can further achieve accurate depth measurement. Therefore, it can effectively improve the detection ability of surface defects.

Microstructure and Mechanical Properties of Heat-affected Zone of Repeated Welding on 304 Stainless Steel

Guo Yanhui, Deng Dong, Sun Zaozhan, Huang Bingchen

2021, 42(4): 198-202. doi: 10.13832/j.jnpe.2021.04.0198

Abstract(390) HTML (120) PDF(31)

Abstract:
The effect of the repeated welding on the microstructure and mechanical properties of the heat-affected zone (HAZ) for 304 stainless steel were investigated under repeated welding up to five times by automatic gas tungsten arc welding using an optical microscope, a x-ray diffraction, a scanning electron microscope and an electron back-scattered diffraction. The repeated welding specimens were consisted with the microstructures of austenitic matrices with lath δ-ferrite. With the increasing of the repeated welding times, the average values of the austenitic grain size increased, and the δ-ferrite content decreased then increased. The preferred orientation of the HAZ changed from <101> to <111>. The values of location misorientation increased monotonically with the increasing number of repeated welding. The variation of the ultimate tensile strength and elongation was affected by the grain size mainly. Due to the hardening rate increasing, the yield tensile strength was increasing.

Defects Analysis of CRDM Weld in EPR Nuclear Power Plants

Tang Jianbang, Yu Zhe, Wang Weiqiang, Sun Jiawei, Lyu Tianming

2021, 42(4): 203-207. doi: 10.13832/j.jnpe.2021.04.0203

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Abstract:
The control rod drive mechanism (CRDM) pressure housing in nuclear power plants is subjected to high temperature, high pressure and high radiation. The pressure housing welds is prone to crack defects, which is the focus of the in-service inspection during the overhaul. In view of the suspected defects found by Ultrasonic inspection in the pre-service inspection, supplement nondestructive test and destructive test were implemented. The results proved that the indications were metallurgical indications caused by the austenitic hardened grain with incomplete recrystallization, which is without effect on the quality of the welds. The methods of signal analysis for suspected defects were summarized.

Multi-Feature Fusion Multi-Step State Prediction of Nuclear Power Sensor Based on LSTM

Zhang Siyuan, Lu Tianyu, Zeng Hui, Xu Chun, Zhang Zhuo, Huang Qingyu, Zhang Yaoyi, Wang Yuanmei

2021, 42(4): 208-213. doi: 10.13832/j.jnpe.2021.04.0208

Abstract(238) HTML (149) PDF(51)

Abstract:
Aiming at the problem in the prediction of nuclear power working condition parameter, this paper uses a large number of time series collected by the nuclear power plant sensor detection system to propose a multi-feature fusion multi-step state prediction model based on long short-term memory network (LSTM). This paper takes the SG1 steam pressure sensor data collected by the real-time parameter system of nuclear power plants as the research object. Firstly, the data is preprocessed for the problems of missing data and inconsistent sampling time scales, and then the structural design and modeling of the multi-feature fusion multi-step state prediction model is completed based on LSTM. Finally, the prediction model proposed in this paper is compared with the multi-step prediction models such as Recurrent Neural Network (RNN), Gated Recurrent Unit (GRU), Model-S1 layer and univariate LSTM. Experimental results show that the fitting performance and prediction performance of the prediction model proposed in this paper are with overall optimization, and it also verifies the applicability of the deep learning method based on the LSTM model in the field of nuclear power plant operation safety assurance.

Study on Wear Mechanism of Thrust Bearing of Nuclear Main Pump in Cooling Water Loss Condition

Cai Long, Wang Weiguang, Lei Mingkai, Li Mengqi, Zhu Bao, Su Xianshun

2021, 42(4): 214-221. doi: 10.13832/j.jnpe.2021.04.0214

Abstract(309) HTML (170) PDF(40)

Abstract:
When the nuclear main pump is in the cooling water loss condition of the nuclear power plant, its thrust bearing loses cold source heat transfer. The temperature of the lubricating medium of the thrust bearing will be continuously increased due to the temperature rise of the bearing, which is accompanied by more complicated thermal transient conditions. When the thickness of the lubricating liquid film of the thrust bearing is seriously reduced, the contact wear of the friction pair occurrs due to the insufficient thickness of the liquid film. A nuclear main pump is disassambled and checked after a water cut-off test, and the wear pattern of the water cut-off operating conditions through lubrication is analyzed. When the station blackout (SBO) inert shutdown is carried out in the water cut-off condition, with the increasing of wear depth, the oil film thickness of the bearing decreases to the extent that it cannot operate reliably, and the loss increases. With the oil film temperature exceeding the Babbitt alloy operating limit temperature of 110-120℃, serious wear of the bearing is prone to occur. It provides a theoretical support for optimizing the bearings and improving the wear resistance of multiple SBO inert shutdowns after wear.

Study on Extension of Containment ILRT Cycle of CPR1000

Fang Xing, Weng Wenqing, Ye Shuixiang, Zhang Wei, Li Jianbo

2021, 42(4): 222-227. doi: 10.13832/j.jnpe.2021.04.0222

Abstract(210) HTML (93) PDF(16)

Abstract:
The containment ILRT for 10 years occupies the critical path of refueling for about 100 hours. 94 nuclear power units in the United States have extended the containment ILRT cycle to 15 years based on containment performance evaluation. This paper introduces the performance evaluation requirements of the relevant containment in the United States in detail, and analyzes the extension of the ILRT cycle by the CPR1000 unit, and the availability of historical test data and inspection data. Taking a CPR1000 demonstration unit as an example, the risk after prolonging the safety test of the containment is calculated, and the risk increment is very small. The results show that the CPR1000 unit basically is with the conditions to extend the test period of the containment seal.

Analysis Model and Parameter Optimization of a Silica Removal System for Boric Acid in Nuclear Power Plant

Xu Liangwang, Liu Bin, Jiang Xiaobin, Tu Zhijian

2021, 42(4): 228-232. doi: 10.13832/j.jnpe.2021.04.0228

Abstract(171) HTML (83) PDF(25)

Abstract:
In order to improve the performance of the silica removal system for boric acid in nuclear power plants, the modified analysis model is established. The simulation of silica removal experiments using 200 Da and 250 Da molecular weight cut off (MWCO) reverse osmosis membranes is carried out by using this model. The calculation results meet the experimental data well, which demonstrates the accuracy and the applicability of the modified model. Further study on parameters optimization of this silica removal system based on this model is implemented as well; the effect of relevant parameters on the performance is analyzed. The conclusion is with reference significance for system design and operation.

Test Analysis of Insulation and Moisture Resistance of Insulation Support Plate of Electrical Penetration in Containment of Nuclear Power Plant

Guo Xing, Chen Qing, Wang Guangjin, Zhou Tian, Qiu Xinyuan, Zhou Yuan, Zhao Yuheng

2021, 42(4): 233-238. doi: 10.13832/j.jnpe.2021.04.0233

Abstract(233) HTML (71) PDF(16)

Abstract:
The variation trend with temperature and relative humidity of water absorption, dielectric loss, relative permittivity, insulation resistance, volume resistivity, surface resistivity, and dielectric strength of polysulfone are measured. Polysulfone used as the insulating support plate on the medium voltage electrical penetration is designed and produced by Nuclear Power Institute of China (NPIC). The result shows that the relative humidity of environment has a significant effect on the water absorption of polysulfone. When the ambient temperature is 23℃ and the relative humidity increases from 30% to 98%, the corresponding water absorption increases from 0.012% to 0.106%, with a growth rate of 783.3%. The insulation resistance decreases gradually with the increasing of ambient temperature and relative humidity, with a maximum decreasing of 99.82%. However, the interphase insulation resistance of the insulation support plate of the electrical penetration is always greater than 200 TΩ. At the same time, the insulation resistance is greatly affected by its surface resistivity. The variation trend of insulation resistance and surface resistivity with ambient temperature and relative humidity is very close. The relative permittivity and dielectric strength are little affected by the ambient temperature and humidity. The insulation support plate of the medium voltage electrical penetration designed and produced by NPIC is with excellent moisture resistance and electrical insulation performance, and can work stably and reliably in high voltage, high temperature and high humidity environment.

Selection and Practice of First of a Kind Test for EPR Unit

Huang Huiming, Yu Weiming, Wu Jiabin

2021, 42(4): 239-243. doi: 10.13832/j.jnpe.2021.04.0239

Abstract(199) HTML (99) PDF(24)

Abstract:
In order to meet the requirements of NNSA, IAEA, NRC, ONR and ASN on the first of a kind test for the new type PWR nuclear power plant, based on the analysis of new design concepts and new design features of the EPR unit and the engineering practice of PWR, this paper proposes a method to determine the first of a kind test by using the control variable selection principle and the five-step selection process. Project practice has proved that this method can ensure the smooth and effective implementation of the commissioning program, and can completely verify the new concept design and the performance of new characteristics, so as to ensure the safe and stable operation of the reactor nuclear power plant. This method can also be applied to other types of PWR nuclear power technology like HPR1000.

Research on Safety Criteria in Design of Floating Nuclear Power Plant

Guo Xiang, Kong Fanfu, Zhu Chenghua, Zhu Gang

2021, 42(4): 244-249. doi: 10.13832/j.jnpe.2021.04.0244

Abstract(182) HTML (151) PDF(64)

Abstract:
As a combination of marine engineering and nuclear power engineering, the floating nuclear power plant is of a new field of nuclear power engineering. There is no corresponding safety design criteria in China. Combined with the actual design requirements of offshore nuclear power platform demonstration project, based on the analysis of the specifications of onshore PWR nuclear power plants, offshore mobile platform and nuclear power ship, the corresponding safety design criteria are proposed from the aspects of overall design, platform and nuclear safety of floating nuclear power plants. The research shows that the safety design of floating nuclear power plants should focus on three basic safety functions.The platform design should consider six factors, including layout, structure, auxiliary system, power, communication and fire protection. The nuclear safety design should fully consider the constraints of island operation and marine application scenarios on the design and operation of nuclear power plant system equipment.

Development of Resin Sampler for Generation Ⅲ Nuclear Power Plants

Zhao Yuheng, Chen Qing, Zhou Tian, Wang Guangjin, Zhou Yuan, Zheng Lanjiang, Qiu Xinyuan

2021, 42(4): 250-253. doi: 10.13832/j.jnpe.2021.04.0250

Abstract(189) HTML (163) PDF(29)

Abstract:
In view of the defects in the traditional way of resin sample in the generation Ⅱ nuclear power plants, this paper introduces a new type of nuclear power plant resin sampler structure and technical characteristics. The pneumatic piston and dynamic seal technology are used in the device, and the remote automatic sampling radioactive resin function is adopted. Test results show that the resin sampler is with high sampling accuracy and good repeatability, and can make the nuclear radioactive waste sampling system more accurate, efficient and safe.

Diagnostic of Loose Parts Event in Water Chamber of Steam Generator

Hu Jianrong, Liu Caixue, Luo Ting, Jian Jie, Yang Taibo, Li Xiang

2021, 42(4): 254-258. doi: 10.13832/j.jnpe.2021.04.0254

Abstract(234) HTML (48) PDF(25)

Abstract:
During the hot performance test in a nuclear power plant, the alarm event of the loose parts of the steam generator occurred. In order to explore the alarm cause for the loose parts and evaluate the damage of the equipment, it is necessary to confirm the loose parts in time. By establishing the diagnosis and analysis model of the loose parts, the original data of the loose event is analyzed in detail, and the propagation speed of the loose signal, the quality and energy of the loose parts are estimated. Finally, the harmfulness of the loose alarm event is evaluated. The results show that the loose part in the steam generator is a moving metal part, which is a loose part of the internal structure of the reactor. The impact frequency of the loose part is 8700~9300 Hz, the peak range of impact acceleration is 3g~90g, the propagation group velocity of the loose signal is 3200~3300 m/s, and the mass of the loose part is estimated to be about 0.1 kg. The impact energy is 0.45~0.89 J. The diagnostic results of mass and energy are in good agreement with the actual situation, which shows that the proposed method is effective.

Numerical Analysis of Aircraft Impact on Wall of High Temperature Gas-Cooled Reactor Building and Evaporator Cavity

Liang Zhenbin, Nie Junfeng, Wang Haitao

2021, 42(4): 259-264. doi: 10.13832/j.jnpe.2021.04.0259

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Abstract:
It is necessary to consider the external event of large aircraft impact on the reactor, which is very important for the safety evaluation of the reactor. In this paper, based on the finite element model of coupled impact dynamics, an equivalent simulation method of double-layer parallel wall subjected to aircraft impact is proposed, and the impact resistance of the thin square evaporator chamber of a high-temperature gas-cooled reactor (HTGR) against commercial aircraft is studied. The external wall of the reactor building is evaluated by the impact of commercial aircraft impact penetration, and the external wall simulation of the commercial aircraft impact reactor building is carried out. Then the residual kinetic energy curve of the aircraft is obtained. The calculation of the aircraft striking the evaporator compartment assumes that the aircraft has no mass loss after passing through the outer wall and is in good shape, striking the square evaporator chamber at the remaining speed. The evaluation shows that the overall damage of the evaporator compartment structure under impact conditions is small and can provide an important barrier function for protecting the critical internal equipment. Assessment results show that the overall damage of evaporator reactor cavity structure under impact condition is small, so the cavity can provide an important barrier function for the important equipment inside it.

Fabrication Design, Qualification and Key Technologies of ITER Gravity Supports

Zhang Teng, Zhang Bo, Wang Yu, Li Pengyuan, He Zujuan, Wei Haihong, Sun Zhenchao, Hou Binglin, Kang Daoan

2021, 42(4): 265-269. doi: 10.13832/j.jnpe.2021.04.0265

Abstract(288) HTML (112) PDF(18)

Abstract:
As a key component of the ITER magnet supports, the ITER gravity support (GS) serves not only to bear the weight of the toroidal field (TF) coil superconductor and the alternate electromagnetic force but also to interrupt the heat from the cryostat ring and to ensure the superconductive state of the TF coils. In this paper, the fabrication design of the GS was simulated and verified via the prototype engineering test. The fatigue performance of the 718 studs was qualified by the 77 K prototype fatigue test. The design of the thermal anchor structure was characterized by the heat exchange test which carried out in a vacuum chamber. Subsequently, based on the fabrication design, the hydraulic tensors and the bespoke high precision bolt elongation measuring device were employed to fasten all the studs accurately. Finally, the results of the leakage test in the vacuum chamber indicated that the leakage rate of GS is much lower than that of ITER requirement. Based on the above work, the fabrication design is feasible and can be utilized in the manufacturing of ITER GS.

Research on Fundamental Characteristics of Nuclear Grade 316H Stainless Steel at Ultra High Temperature

Zhang Hongliang, Zhu Mingdong, Sun Xiaoyang, He Daming, Wang Qingtian, Su Dongchuan, Li Ning, Zeng Chang, He Xikou

2021, 42(4): 270-276. doi: 10.13832/j.jnpe.2021.04.0270

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Abstract:
A fundamental feature of the fourth-generation-reactor is that most of the designed operating temperature are between 500℃ to 800℃, while the traditional material system and data of the PWR are below 350℃, which cannot meet the requirements. In this paper, 316H is selected as the research object through demonstration and analysis, which is suitable for most reactors and closest to engineering application. The experimental study on mechanical properties, specific heat capacity, average linear expansion coefficient, intergranular corrosion characteristics and low cycle fatigue at 800℃ was carried out. the result shows that the measured data are significantly higher than the standard values. It is recommented that the temperature limit for the long-term operation shall not exceed 700℃, and the temperature limit for the short-term operation shall not exceed 800℃. This study provides a basis for the selection and evaluation of the structural materials of the fourth-generation-reactor.

Analysis of Blockage Accident of Lead-Based Fast Reactor Single-Box Fuel Assembly Based on CFD

Chen Baowen, Deng Jian, Ling Yufan, Hu Baolong, Wang Tianshi, Zhu Enping, Wang Ting

2021, 42(4): 277-281. doi: 10.13832/j.jnpe.2021.04.0277

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Abstract:
The corrosion products produced during the operation of the lead-based reactor may be deposited in the reactor, resulting in the occurrence of flow plugging accident. Based on the computational fluid dynamics software Ansys Fluent, the effects of different blockage area, thickness, type and position on the deterioration of heat transfer and the properties of flow field in current plugging accident are analyzed. The results show that the increasing of blockage area will increase the reflux area, which makes the temperature fall more slowly and the heat transfer worsen obviously; the increasing of the blockage thickness will lead to the increasing of the maximum temperature of coolant and cladding, which can easily lead to cladding damage; the coolant in the porous medium blockage passes at a lower flow rate, which alleviates the influence of blockage and does less harm than the solid blockage. The local warming caused by the blockage located in the middle of the active area is more obvious and more harmful than that caused by the blockage in the entrance and exit of the active area.

Study on Fast Fracture Evaluation Method for Reactor Pressure Vessel with Excessive Carbon Content

Su Dongchuan, Xie Hai, Zhang Yixiong, Cui Huaiming, Wu Lin

2021, 42(4): 282-288. doi: 10.13832/j.jnpe.2021.04.0282

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Abstract:
The initial non-ductile transition temperature is an important input parameter in fast fracture evaluation according to the RCC-M code. The initial non-ductile transition temperature (RT_NDT) will be affected by the carbon content, but there is no quantitative relationship between them. When the carbon content of the reactor pressure vessel exceeds the standard, it is necessary to complete the fast fracture evaluation without quantitative relationship to ensure the integrality of RPV. In this paper, the evaluation method of the fast fracture of RPV with excessive carbon content is studied. Taking the RPV core barrel with excessive carbon content as an example, the defect repair is considered, and the fast fracture analysis and evaluation are carried out by calculating the maximum initial non-ductile transition temperature. This method can support the operation and the in-service inspection of the pressure vessels with excessive carbon content.

2021 Vol. 42, No. 4