28 results on '"Chen, Jingen"'
Search Results
2. Minor Actinides Transmutation in Thermal, Epithermal and Fast Molten Salt Reactors with Very Deep Burnup
- Author
-
Zou, Chunyan, Yu, Chenggang, Zhou, Jun, Chen, Shuning, Wu, Jianhui, Zou, Yang, Cai, Xiangzhou, Chen, Jingen, and Liu, Chengmin, editor
- Published
- 2023
- Full Text
- View/download PDF
3. Effect of 135Xe and 135Xem Migration on Reactivity in Molten Salt Reactor
- Author
-
WU Chen;YU Chenggang;CAI Xiangzhou;CHEN Jingen
- Subjects
molten salt reactor ,xenon poison ,135xem ,helium bubble system ,Nuclear engineering. Atomic power ,TK9001-9401 ,Nuclear and particle physics. Atomic energy. Radioactivity ,QC770-798 - Abstract
Fissile material dissolved in the molten salt is employed in molten salt reactor(MSR), which circulates through the primary loop. Considering the flow characteristic of fuel in MSR, the neutron poison calculation model such as Xe is different from traditional solid reactor. As an important neutron poison, 135Xe is free to migrate with the flow of molten salt in MSR. The helium bubble system of molten salt reactor can blow the fission products of krypton, xenon and other gases out of the reactor core to improve the neutron economy. The recent TENDL2021 nuclear data library estimates the thermal neutron absorption cross section of 135Xem, which is much higher than 135Xe thermal neutron absorption cross section. In order to predict the effect of xenon on core reactivity more accurately, the effect of 135Xem thermal neutron absorption cross section was added into the Xe poison model based on the lumped volume method. The migration model is made up of cover gas calculation model and xenon worth calculation model. In order to calculate the xenon worth, the loop void fraction was calculated firstly, and the loop void fraction was inserted into the xenon worth calculation model. According to the rate balance equations and mass transfer equations, the concentration of xenon in loop could be calculated to compute the xenon worth. Firstly, the model results considering the flow effect of fuel salt were verified with the experimental results of 8 MW molten salt reactor experiment (MSRE). As the results shown, with the increase of loop void fraction, the steadystate xenon worth increases first and decreases later, which is in good agreement with experimental results of MSRE. Then the influence of fuel salt types and 135Xem on steadystate xenon worth was considered in this paper. The steadystate xenon worth of the 233U system with and without 135Xem is made and compared to a 235U system, indicating that the influence of 135Xem on xenon worth cannot be ignored, which is much higher with the increase of power. Finally, the contribution of graphite, fuel salt and helium bubble to Xe worth was evaluated by the interaction between graphite and helium bubble, indicating that the interaction between helium bubble and graphite cannot be ignored.
- Published
- 2022
4. Simulation of Physical Parameters for a Photoneutron Source
- Author
-
Wang, Xiaohe, Liu, Longxiang, Hu, Jifeng, Han, Jianlong, Yang, Pu, Lin, Zuokang, Zhang, Guilin, Wang, Naxiu, Cai, Xianzhou, Wang, Hongwei, and Chen, Jingen
- Subjects
Physics - Computational Physics ,Physics - Instrumentation and Detectors - Abstract
A compact photoneutron source (PNS), based on an electron linac was designed and constructed to provide required nuclear data for the design of Thorium Molten Salt Reactor (TMSR). Many local shielding are built to reduce the background of neutron and {\gamma} rays, making the location of the time of flight (TOF) detector be fixed at 6.2 m place. Under the existing layout, some physical parameters are very difficult to get by the experiments, while can be obtained by the Monte Carlo simulation method. However, for the deep penetration problem of the neutron and {\gamma} rays transport in the channel of PNS with complex geometry, the normal Monte Carlo method is inefficient since electron transport calculation need a large amount of computing time and neutrons have little contribution to the detector in far-source region. In this work, the subsection method is applied in the simulation for PNS, which divide the simulation process in two steps, recording the neutron and {\gamma} rays information passing through the source window in the first step and adopting the covariance reduction techniques in the second step. The simulated neutron flux and energy spectrum at the TOF detector place with the relative error 1.6% are well agreement with the experimental results, achieving an efficiency 23 times better than the normal method. This method is fast and efficient in predicting the physical parameters, providing a required verification and initiating the foreseen physics experiment., Comment: 10 pages, 7 figures, 2 tables
- Published
- 2017
5. Measurements of the Total Cross section for Thermal Neutrons at PNS
- Author
-
Liu, Longxiang, Wang, Hongwei, Ma, Yugang, Cao, Xiguang, Cai, Xiangzhou, Chen, Jingen, Zhang, Guilin, Han, Jianlong, Zhang, Guoqiang, Hu, Jifeng, Wang, Xiaohe, Li, Wenjiang, Yan, Zhe, and Fu, Haijuan
- Subjects
Nuclear Experiment ,Physics - Instrumentation and Detectors - Abstract
In order to measure the total cross section for thermal neutrons, a photoneutron source (PNS, phase 1) has been developed for the acquisition of nuclear data for the Thorium Molten Salt Reactor (TMSR) at the Shanghai Institute of Applied Physics (SINAP). PNS is an electron LINAC pulsed neutron facility that uses the time-of-flight (TOF) technique. It records the neutron TOF and identifies neutrons and $\gamma$-rays by using a digital signal processing technique. The background is obtained by using a combination of employing 12.8 cm boron-loaded polyethylene(PEB) (5$\%$ w.t.) to block the flight path and Monte Carlo methods. The neutron total cross sections of natural beryllium are measured in the neutron energy region from 0.007 to 0.1 eV. The present measurement result is compared with the fold Harvey data with the response function of PNS., Comment: 9 pages,10 figures
- Published
- 2017
6. Measurement of Neutron Total Cross Section of Th in 0.008-0.1 eV at TMSR-PNS
- Author
-
HU Jifeng;WANG Xiaohe;JIANG Bing;HAN Jianlong;CHEN Jingen;CAI Xiangzhou
- Subjects
tmsr-pns ,total cross section ,tho2 ,transmission method ,Nuclear engineering. Atomic power ,TK9001-9401 ,Nuclear and particle physics. Atomic energy. Radioactivity ,QC770-798 - Abstract
The nuclear data of Th is primary interest for the thorium-uranium (Th-U) fuel cycle, and its nuclear data accuracy is related to the conversion performance of Th-U fuel cycle. Based on the photo-neutron source for the thorium molten salt reactor (TMSR-PNS), the total cross section (TCS) of ThO2 was measured by the transmission method for the high purity thorium oxide (ThO2) samples. ThO2 samples which the thicknesses were respectively 0.9 cm and 0.608 cm were made by pressing and sintering process at 1 800 ℃. Based on the different pulse shapes of neutrons and γ-rays (n/γ) signal from the lithium iodide (LiI) scintillator detector used for the time-of-flight (TOF) calculation, γ-rays were identified by pulse shape discrimination (PSD) method to reduce the interference of the measured cross section. Combined with the trough position of TOF spectrum and its corresponding energy for the calibration targets of Ag, In and Cd, the flight distance was fitted to be 5.899 m, so as to convert TOF spectrum of neutrons into energy spectrum. Energy spectrum of 10 cm thickness boracic polythene sample was measured to deduct the influence of the scattering neutron from the background. According to the measured neutron leakage spectrum with or without ThO2 sample, experimental data of TCS for ThO2 were obtained. TCS of 16O from ENDF/B-Ⅶ.1 was deducted from TCS of ThO2 and TCS of Th was obtained. Combined with error of the sample purity and thickness, the fitting flight distance, relative standard deviation of TCS for 16O from ENDF/B-Ⅶ.1, and experimental statistical error, the uncertainty of experimental data were calculeted to be 3.25%-4.51% in the energy range from 0.02 to 0.1 eV. The results show that the measured data of TCS of Th are good agreement with the evaluated data of ENDF/B-Ⅶ.1 in the energy range from 0.02 to 0.1 eV. In the thermal energy range (E
- Published
- 2022
- Full Text
- View/download PDF
7. Core design optimization for a novel heavy water moderated molten salt reactor
- Author
-
WU Jianhui, YU Chenggang, ZOU Chunyan, MA Yuwen, JIA Guobin, CAI Xiangzhou, and CHEN Jingen
- Subjects
hwmsr ,thermal hydraulic ,neutronics ,thorium-uranium fuel cycle ,Nuclear engineering. Atomic power ,TK9001-9401 - Abstract
BackgroundHeavy water moderated molten salt reactor (HWMSR) adopts heavy water as moderator and heavy metal elements dissolved in fluoride salt as the fuel, which has high neutron economy. But large temperature difference of fuel salt outlet will lead to the thermal fatigue of piping components at the top of the core.PurposeThis paper aims to optimize the core design of HWMSR to minimize the fuel salt outlet temperature.MethodsBy using the developed codes of neutron-thermal hydraulic coupling calculation and core critical search calculation, the power density distribution, outlet temperature distribution, initial 233U loading and breading ratio (BR) were analyzed for the cores with different size of molten salt channels.ResultsThe calculation results demonstrate that increasing the radius of molten salt channels in the inner core zone (correspondingly decreasing the radius of molten salt channels in the outer core zone) will decrease the peak of power density and the maximum outlet temperature of molten salt while the influence on BR and 233U initial loading.ConclusionsThis study provides a valuable foundation for core optimization of HWMSR.
- Published
- 2021
- Full Text
- View/download PDF
8. Influence of thermal neutron scattering effect of FLiBe molten salt on neutronic performance of molten salt reactors
- Author
-
ZHANG Zhicheng, HU Jifeng, CHEN Jingen, and CAI Xiangzhou
- Subjects
molten salt reactor ,flibe ,thermal neutron scattering cross-section ,effective multiplication factor ,Nuclear engineering. Atomic power ,TK9001-9401 - Abstract
BackgroundFLiBe is commonly used as the coolant and carrier salt in liquid molten salt reactors (MSRs). Its certain moderating properties and thermal neutron scattering attributes affect the neutronic performance of the MSR, and this in turn influences the physical design and safe operation of the reactor. Consequently, studying FLiBe's thermal neutron scattering data is essential for MSRs.PurposeThis study aims to analyze the influence of of FLiBe thermal neutron scattering on neutronic performances of a 65-MW MSR.MethodsFirst, according to the requirements, a core model of a 65-MW MSR was established by using the general Monte Carlo procedure. Then, the neutronics performance of the MSR was calculated by considering the scattering cross-section of the free gas model and FLiBe thermal neutron scattering data (e.g., the neutron spectrum, effective multiplication factor, and nuclide reactivity rate). Finally, the changes in the influence of FLiBe thermal neutron scattering effect on neutronic properties under different energy spectra were compared.ResultsThe computation results show that, by considering the thermal scattering effect of FLiBe molten salt, the neutron energy spectrum in the core of the MSR becomes harder, 235U fission rate decreases, the keff value of the reactor decreases, but the density coefficient in the temperature reaction coefficient of the fuel keeps almost unchanged, and the Doppler coefficient decreases by 0.28×10-5 K-1. With the hardening of the energy spectrum, the variation in the 235U fission rate reduction decreases, and the decrease in keff caused by thermal neutron scattering changs from 9.2×10-4 to 2×10-4.ConclusionsTherefore, it is necessary to incorporate FliBe's thermal neutron scattering data into the physical calculations for the MSR core.
- Published
- 2023
- Full Text
- View/download PDF
9. TOF spectroscopy measurement using waveform digitizer
- Author
-
Liu, Longxiang, Wang, Hongwei, Ma, Yugang, Cao, Xiguang, Cai, Xiangzhou, Chen, Jingen, Zhang, Guilin, Han, Jianlong, Zhang, Guogiang, Hu, Jifeng, and Wang, Xiaohe
- Subjects
Physics - Instrumentation and Detectors ,Nuclear Experiment - Abstract
The photoneutron source (PNS, phase 1), an electron linear accelerator (linac)-based pulsed neutron facility that uses the time-of-flight (TOF) technique, was constructed for the acquisition of nuclear data from the thorium molten salt reactor(TMSR) at the Shanghai Institute of Applied Physics (SINAP). The neutron detector signal, with the information on the pulse arrival time, pulse shape, and pulse height, was recorded by using a waveform digitizer (WFD). By using the pulse height and pulse-shape discrimination (PSD) analysis to identify neutrons and $\gamma$-rays, the neutron TOF spectrum was obtained by employing a simple electronic design, and a new WFD-based DAQ system was developed and tested in this commissioning experiment. The developed DAQ system is characterized by a very high efficiency with respect to millisecond neutron TOF spectroscopy, Comment: 5 pages, 7 figures
- Published
- 2015
- Full Text
- View/download PDF
10. Conceptual Design of a Novel Megawatt Molten Salt Reactor Cooled by He-Xe Gas
- Author
-
Zhao, Hongkai, primary, Wu, Jianhui, additional, Chen, Shuning, additional, Cui, Yong, additional, Chen, Jingen, additional, and Cai, Xiangzhou, additional
- Published
- 2023
- Full Text
- View/download PDF
11. Flow and heat transfer characteristics of FLiNaK salt in an annular channel heater
- Author
-
YANG Yang, ZOU Yang, CHEN Jingen, and ZOU Chong
- Subjects
wire coil ,annular channel heater ,flinak salt ,flow and heat transfer characteristics ,energy efficiency ,Nuclear engineering. Atomic power ,TK9001-9401 - Abstract
BackgroundThe simulation device is one-third scale of the 10 MW solid-fuel thorium molten-salt reactor (TMSR) and has the rated heat power of 370 kW. Nineteen annular channel heaters with wire coil were designed to heat the simulation device to the desired temperature.PurposeThis study aims to further investigate the effect of outer diameter of heating tube and annular width on the flow and heat transfer characteristics of FLiNaK salt in the annular channel heater.MethodsThe flow and heat transfer characteristics of annular channel heater with wire coil were simulated by using Fluent software. The effects of outer diameter of heating tube on the flow and temperature fields were investigated numerically, so did the effects of annular width on the flow and heat transfer characteristics. The overall performance of annular channel heater was evaluated by the principle of entropy production.ResultsThe simulation results show that the axial velocity, radial velocity and tangential velocity are generated and meanwhile counter-flow is formed on longitudinal section.ConclusionsThe outer diameter of heating tube has no effect on the flow and heat transfer characteristics. The larger the outer diameter of heating tube, the higher the energy efficiency. As the increase of the annular width, the pressure loss, the heat transfer coefficient decreases, and the energy efficiency gradually decreases.
- Published
- 2022
- Full Text
- View/download PDF
12. Accident occurrence frequency of a small helium-xenon gas cooled nuclear reactor system
- Author
-
WU Jianhui, ZHOU Jun, ZOU Chunyan, JIA Guobin, ZHANG Ao, CAI Xiangzhou, and CHEN Jingen
- Subjects
small helium-xenon cooled reactor ,probabilistic safety assessment ,fault tree ,Nuclear engineering. Atomic power ,TK9001-9401 - Abstract
BackgroundHelium-xenon gas cooled nuclear reactor adopts helium-xenon gas as the coolant and is able to apply the integral Brayton cycle. It has advantages in miniaturization and lightweight design, which has attracted more and more attention worldwide. But rare severe accidents studies have been carried out for the small helium-xenon gas cooled reactor. Probabilistic safety assessment (PSA) is an important method to evaluate the safety of a reactor system. The results obtained by PSA can provide a valuable basis for improving the core design, identifying fault, guiding operation, etc. whilst the occurrence frequency of initial events is required for PSA analysis.PurposeThis paper aims to evaluate the accident occurrence frequency for the small helium-xenon gas cooled reactor to provide an input for PSA analysis.MethodsBased on the main technical characteristics of a small innovation helium-xenon cooled mobile nuclear power system (SIMONS), the occurrence frequency of typical accidents was analyzed by referring to the operational experiences and data of high temperature gas cooled reactor (HTGR) and light water reactor (LWR). Intensity analyses were performed on the frequency of accidents such as increase or decrease of core heat removal, abnormal reactivity and power distribution, pipe break and abnormal equipment leakage, anticipated transient without scram (ATWS), and loss of offsite power (LOOP).ResultsThe calculation results show that the accidents of heat removal increase and decrease, abnormal reactivity and power distribution, pipe break and abnormal equipment leakage, ATWS, and LOOP have an occurrence frequency of 3.90×10-2 RY-1, 2.36×10-1 RY-1, 2.69×10-2 RY-1, 6.50×10-2 RY-1, 2.69×10-2 RY-1 and 4.60×10-2 RY-1, respectively.ConclusionsThe calculated results can be taken as the inputs for the PSA study of the SIMONS, providing basic reference value for further PSA analysis of mobile nuclear power system.
- Published
- 2022
- Full Text
- View/download PDF
13. Design of beam shaping assembly for boron neutron capture therapy based on D-T neutron source
- Author
-
ZHU Yinan, LIN Zuokang, LU Linyuan, YU Haiyan, CHEN Jingen, and XIE Leidong
- Subjects
boron neutron capture therapy ,d-t neutron source ,monte carlo simulation ,beam shaping assembly ,Nuclear engineering. Atomic power ,TK9001-9401 - Abstract
BackgroundBoron neutron capture therapy (BNCT) is a promising treatment for cancer. In recent years, D-T neutron source has shown great potential as a suitable neutron source for BNCT. However, the neutron energy produced by a D-T neutron source is 14.1 MeV, which cannot be straightforwardly used in a BNCT, so beam shaping is necessary.PurposeThis study aims to design beam shaping assembly (BSA) to make the neutron beam of D-T neutron source suitable for BNCT.MethodsThe Monte Carlo simulation program MCNP5 was employed to design the corresponding BSA for D-T neutron source. The advantages of using natural uranium as neutron multiplicative layer was firstly verified by simulation. Then, the moderating layer was simulated together the supplementary moderating layer, reflective layer, shielding layer and thermal neutron absorption layer using different materials and thickness to obtain optimal configuration of these materials.ResultsSimulation results show that an ideal therapeutic neutron beam is obtained by using a 14 cm radius natural uranium as neutron multiplicative, 50 cm thick BiF3 and 10 cm thick TiF3 as a double-moderating layer, a 17 cm thick AlF3 as a supplementary moderator, a 0.2 mm thick Cd as a thermal neutron absorbing layer, a 3.5 cm thick Pb as a γ shielding layer and a 10 cm thick Pb as a reflecting layer. Combination of these materials is suggested to optimize and design as the neutron BSA.ConclusionsThe air outlet parameters of output neutron beam using BSA designed in this study can well meet the recommended values of the International Atomic Energy Agency (IAEA).
- Published
- 2022
- Full Text
- View/download PDF
14. Core preliminary neutronics design of a martian surface molten salt reactor with 1 MWth
- Author
-
HU Guang, CUI Deyang, LU Linyuan, LI Xiaoxiao, CHEN Jingen, and CAI Xiangzhou
- Subjects
molten salt reactor ,liquid metal heat pipe ,control drum ,Nuclear engineering. Atomic power ,TK9001-9401 - Abstract
BackgroundNuclear reactors have become the main energy supply for future manned Mars exploration missions due to their high energy density, high power-mass ratio and small size. Heat pipe molten salt reactor is an innovative concept combining molten salt reactor and high temperature heat pipe technology.PurposeThis study aims to carry out preliminary neutronics design of a martian surface molten salt reactor (MS)2R core with 1 MWth and a lifetime longer than 5 a.MethodsFirst of all, the core model of (MS)2R was estabished according to the requirements. It mainly included active zone (including fuel salt and heat pipe), active zone wall, reflector, reactor vessel, etc. Then, the MCNP (Monte Carlo N particle Transport Code) and MOBAT were used to optimize the core size and reactivity control of (MS)2R.ResultsThe physical design parameters of the core are obtained: the height and diameter of the core are 90.94 cm and 88.94 cm, respectively, the uranium inventory is 146.08 kg, and the core mass is 2.09×103 kg. Reactivity control is achieved by a control drum system, in which the B4C (10B enrichment is 90%) with a thickness of 1.8 cm and a wrap angle of 120° is used as the neutron absorber.ConclusionsThe control drum arrangement can meet the critical safety requirements during the lifecycle of (MS)2R under full power (1 MWth) operation for five years, and the number of heat pipes can meet the heat transfer safety limit. This study provides a basic theoretical reference for the design of planet surface molten salt reactor.
- Published
- 2021
- Full Text
- View/download PDF
15. Research on Initiating Events Analysis of Small Helium-Xenon Gas Cooled Nuclear Reactor.
- Author
-
Zhou, Jun, Wu, Jianhui, Cui, Yong, Zhao, Hongkai, Zou, Chunyan, and Chen, Jingen
- Subjects
GAS cooled reactors ,NUCLEAR reactors ,MOLTEN salt reactors ,PRESSURIZED water reactors ,NUCLEAR energy ,CONCEPTUAL design - Abstract
Initiating event analysis is an essential prerequisite of conducting probabilistic safety assessment for nuclear reactors, which plays an important role in improving the core design, identifying fault, and guiding operation. In order to determine the initiating event list of SIMONS (Small Innovative helium-xenon cooled Mobile Nuclear power System), preliminary researches on the initial event of SIMONS were carried out using the MLD (Main Logic Diagram) analysis method and referring to the initial event list and initial event analysis theory of other nuclear reactors such as HTGR (High Temperature Gas-cooled Reactor), MSR (Molten Salt Reactor), and PWR (Pressurized water reactor). With employing these methods, a total of 31 initial events are identified for SIMONS based on its latest conceptual design. These initial events are then divided into six groups according to the accident types, which are core heat removal increase, core heat removal decrease, abnormal reactivity and power distribution, pipeline crevasse and equipment leakage, anticipated transients without scram, and disasters (internal and external). The obtained results can provide a theoretical basis for the further safety analysis of SIMONS. [ABSTRACT FROM AUTHOR]
- Published
- 2023
- Full Text
- View/download PDF
16. A Review of Molten Salt Reactor Multi-Physics Coupling Models and Development Prospects
- Author
-
Wu, Jianhui, primary, Chen, Jingen, additional, Cai, Xiangzhou, additional, Zou, Chunyan, additional, Yu, Chenggang, additional, Cui, Yong, additional, Zhang, Ao, additional, and Zhao, Hongkai, additional
- Published
- 2022
- Full Text
- View/download PDF
17. Design Analysis of a Helium Xenon-Printed Circuit Heat Exchanger for a Closed Brayton Cycle Microtransport Reactor
- Author
-
Yin, Hui, Ma, Yuwen, Jia, Guobin, and Chen, Jingen
- Abstract
Microtransport with small size and a wide range of applications is very attractive for the utilization of future reactor modularization. A microreactor uses a closed Brayton cycle (CBC) to achieve high conversion efficiencies at low specific mass. The recuperator is one of the key components of CBC system which recovers heat exhausted from the turbine. A printed circuit heat exchanger (PCHE) is currently the preferred type of recuperator for the closed Brayton cycle due to its high heat transfer efficiency, high compactness, and high pressure and temperature resistance. This work intends to analyze different geometry configurations of a zigzag PCHE to increase its heat transfer efficiency while reducing its weight and size. The effect of geometric structure parameters such as channel diameter, zigzag pitch length, and zigzag angle on the Nusselt number and Fanning friction factor is investigated using a zigzag PCHE unit model. Based on the numerical simulation, the principle of least squares is employed to carry out a nonlinear fitting of the flow and heat transfer criterion correlation equations. Besides, the maximum value of the Nusselt number and minimum value of the Fanning friction factor are optimized as two conflicting objective functions using the nondominated sorting genetic algorithm-II (NSGA-II), with which a set of optimal solutions is obtained. Meanwhile, the shortest normalized distance is used to determine the compromised solution on the Pareto optimal points, and the independent variable’s sensitivity analysis is performed. Finally, a multiobjective optimization analysis is conducted for the PCHE to achieve lightweight and the high heat transfer efficiency design requirements of SIMONS.
- Published
- 2023
- Full Text
- View/download PDF
18. Accident Modeling and Analysis of Nuclear Reactors
- Author
-
Wu, Jianhui, primary, Chen, Jingen, additional, Zou, Chunyan, additional, and Li, Xiaoxiao, additional
- Published
- 2022
- Full Text
- View/download PDF
19. The Conceptual Design of Electron-accelerator-driven Subcritical Thorium Molten Salt System
- Author
-
Lin, Zuokang, Chen, Jingen, Guo, Wei, and Dai, Zhimin
- Published
- 2013
- Full Text
- View/download PDF
20. Transient release of radioactive iodine from the fission of UF4 in 2LiF–BeF2 salt
- Author
-
Geng, Junxia, primary, Zhao, Zhongqi, additional, Cheng, Zhiqiang, additional, Li, Wenxin, additional, Dou, Qiang, additional, Fu, Haiying, additional, Hu, Jifeng, additional, Cai, Xiangzhou, additional, Chen, Jingen, additional, and Li, Qingnuan, additional
- Published
- 2021
- Full Text
- View/download PDF
21. Evaporation behavior of 2LiF–BeF2–ZrF4 molten salt with irradiated nuclear fuel
- Author
-
Zhou, Jinhao, primary, Geng, Junxia, additional, Luo, Yan, additional, Cui, Rongrong, additional, Zhao, Zhongqi, additional, Fu, Haiying, additional, Dou, Qiang, additional, Wang, Xiaohe, additional, Li, Wenxin, additional, Chen, Jingen, additional, and Li, Qingnuan, additional
- Published
- 2021
- Full Text
- View/download PDF
22. Distribution and behaviour of 233Pa in 2LiF–BeF2 molten salt
- Author
-
Zhao, Zhongqi, primary, Hu, Jifeng, additional, Cheng, Zhiqiang, additional, Geng, Junxia, additional, Li, Wenxin, additional, Dou, Qiang, additional, Chen, Jingen, additional, Li, Qingnuan, additional, and Cai, Xiangzhou, additional
- Published
- 2021
- Full Text
- View/download PDF
23. Evaporation behavior of 2LiF–BeF2–ZrF4 molten salt with irradiated nuclear fuel.
- Author
-
Zhou, Jinhao, Geng, Junxia, Luo, Yan, Cui, Rongrong, Zhao, Zhongqi, Fu, Haiying, Dou, Qiang, Wang, Xiaohe, Li, Wenxin, Chen, Jingen, and Li, Qingnuan
- Published
- 2021
- Full Text
- View/download PDF
24. Transient release of radioactive iodine from the fission of UF4 in 2LiF–BeF2 salt.
- Author
-
Geng, Junxia, Zhao, Zhongqi, Cheng, Zhiqiang, Li, Wenxin, Dou, Qiang, Fu, Haiying, Hu, Jifeng, Cai, Xiangzhou, Chen, Jingen, and Li, Qingnuan
- Published
- 2021
- Full Text
- View/download PDF
25. Distribution and behaviour of 233Pa in 2LiF–BeF2 molten salt.
- Author
-
Zhao, Zhongqi, Hu, Jifeng, Cheng, Zhiqiang, Geng, Junxia, Li, Wenxin, Dou, Qiang, Chen, Jingen, Li, Qingnuan, and Cai, Xiangzhou
- Published
- 2021
- Full Text
- View/download PDF
26. Evaporation behavior of 2LiF-BeF 2 -ZrF 4 molten salt with irradiated nuclear fuel.
- Author
-
Zhou J, Geng J, Luo Y, Cui R, Zhao Z, Fu H, Dou Q, Wang X, Li W, Chen J, and Li Q
- Abstract
The evaporation behaviours of various components were investigated by using a low pressure distillation method in a 2LiF-BeF
2 -ZrF4 mixture containing irradiated ThF4 and UF4 . The experiment showed that BeF2 and ZrF4 were found to mainly condensate at the outer cover, the coolest zone, and their relative volatilities vs. LiF were 9.8 and 32.2, respectively, while for ThF4 and UF4, at four different temperature zones the values were almost constant, at 0.1 and 0.3. The radioactivity of various nuclides was further detected using gamma spectrometer analysis.137 Cs was hardly observed due to long half-time decay.233 Pa was found to co-evaporate with the carrier salt, while239 Np mainly remained in the residual salt as237 U. In different temperature zones, the decontamination factors of rare earth in receiver salts ranged from 10 to 103 . On the basis of the investigation, it was proposed that the distribution of various nuclides after distillation, may be helpful to design the feasible condensate system to recover the carried salt in a molten salt reactor., Competing Interests: There are no conflicts to declare., (This journal is © The Royal Society of Chemistry.)- Published
- 2021
- Full Text
- View/download PDF
27. Transient release of radioactive iodine from the fission of UF 4 in 2LiF-BeF 2 salt.
- Author
-
Geng J, Zhao Z, Cheng Z, Li W, Dou Q, Fu H, Hu J, Cai X, Chen J, and Li Q
- Abstract
In this study, the behavior of fission product iodine released from the melting process of a mixture consisting of UF
4 irradiated with neutrons and 2LiF-BeF2 (FLiBe) salt was studied. The experiment showed that a large amount of iodine was released immediately during melting and captured by Ni metal foils. The transient release of iodine observed in this experiment is attributed to the redox reaction between the hot atoms of the fission product iodine that cumulated due to long-time irradiation. The effect of the redox status of the molten salt on the transient release of iodine was also investigated. Based on this investigation, it was proposed that the activity ratios of131 I to salt-seeking fission products in the fuel salt, as an effective diagnostic criterion, may be used for the surveillance of the redox potential of fuel salts in a molten salt reactor., Competing Interests: There are no conflicts to declare., (This journal is © The Royal Society of Chemistry.)- Published
- 2021
- Full Text
- View/download PDF
28. Distribution and behaviour of 233 Pa in 2LiF-BeF 2 molten salt.
- Author
-
Zhao Z, Hu J, Cheng Z, Geng J, Li W, Dou Q, Chen J, Li Q, and Cai X
- Abstract
Distribution and behavior of
233 Pa, essential in the thorium-uranium nuclear fuel cycle, were studied in 2LiF-BeF2 (66 : 34 mole%, FLiBe) molten salt by γ-ray spectrometry. The experiments showed that233 Pa deposited slightly on the surface of graphite crucible. The addition of Hastelloy and metallic lithium decreased the233 Pa specific activity in the salt by 1 to 2 orders of magnitude rapidly. Analysis indicated that reductive deposition of233 Pa was responsible for the rapid decrease of233 Pa specific activity in the salt. Additional experiments strongly supported the mechanism of reductive deposition of233 Pa induced by Hastelloy and metallic lithium. In view of the large deposition of233 Pa on Hastelloy, the possible influence of fissile nuclide233 U produced from233 Pa decay on the operation of thorium-based molten salt reactor was discussed., Competing Interests: There are no conflicts to declare., (This journal is © The Royal Society of Chemistry.)- Published
- 2021
- Full Text
- View/download PDF
Catalog
Discovery Service for Jio Institute Digital Library
For full access to our library's resources, please sign in.