核动力工程

Research on Efficiency of 2D Ray Tracing under Different Geometric Descriptions

Xu Fei, Luo Qi, Ming Pingzhou, Yao Dong, Huang Wei, Yu Hongxing

2018, 39(S2): 1-5. doi: 10.13832/j.jnpe.2018.S2.0001

Abstract(19) PDF(0)

Abstract:
In order to develop an autonomous geometric preprocessing module and, moreover, adapt to the future large-scale parallel computing, the content of ray tracing efficiency is studied. Based on the constructive solid geometry, CSG, and the point-line geometry, which are implemented respectively in OpenMOC and KYLIN-2, the serial tracing efficiencies under these two geometric descriptions are compared. The ray tracing procedure of KYLIN-2 is parallelized in space considering the physical meaning. The results show that with the help of Computer Aided Design, CAD, modeling tool and parallel ray tracing in space, the efficiency of ray tracing has been obviously improved under the point-line geometry. Parallel ray tracing in space is feasible and achieves parallel scalability with high speedup ratio.

Steady-State Lattice Boltzmann Method for Neutron Transport with GPU Acceleration

Ma Yu, Wang Yahui, Peng Xingjie, Xia Bangyang

2018, 39(S2): 6-9. doi: 10.13832/j.jnpe.2018.S2.0006

Abstract(12) PDF(0)

Abstract:
The multi-dimensional neutron transport process is simulated by using the lattice Boltzmann method（LBM） with strong localization. Meanwhile, to improve the computational speed of LBM, the Graphics Processing Unit（GPU） acceleration is applied to speed up the LBM calculation in parallel. The numerical solutions for typical neutron transport cases show that the LBM can simulate the neutron transport problem accurately and the GPU acceleration can efficiently improve the computational speed of the LBM. The combination of these two techniques can realize the efficient and accurate neutron transport calculation.

Research on Top Peak Flux Problem of Mixed Core with MOX Fuel

Li Songling, Li Tianya, Guo Xingkun, Yu Yingrui, Li Qing

2018, 39(S2): 10-14. doi: 10.13832/j.jnpe.2018.S2.0010

Abstract(16) PDF(2)

Abstract:
Through detailed modeling and analysis, this paper expounds the phenomenon and reason for the top peak flux. It is found that the top peak flux is widespread in the mixed core with MOX fuel. In order to solve this problem, it is proposed to change the top composition of MOX assembly or UO₂ assembly to reduce the top neutron flux. Calculations show that adding 12～18 cm Gd₂O₃ to the top of the MOX fuel rod is the best solution. Center drilling of MOX fuel can also be another solution.

Sensitivity Analysis Based on 2D/1D Transport Code KYCORE

Wu Qu, Peng Xingjie, Tang Xiao, Shi Guanlin, Yu Yingrui, Li Qing, Wang Kan

2018, 39(S2): 15-19. doi: 10.13832/j.jnpe.2018.S2.0015

Abstract(9) PDF(0)

Abstract:
Adjoint sensitivity analysis（ASA） and the forward sensitivity analysis（FSA） based on the reduced order module（ROM） were developed on the 2D/1D transport code KYCORE. Sensitivity analysis on the TMI-1 PWR cell benchmark was performed. Results show that sensitivity coefficients of the effective multiplication factor（keff） with respect to 235 U fission and capture cross sections calculated by KYCORE and RMC agree well with each other. However, the dependency exists in the resonance region. Sensitivity coefficients calculated by the FSA based on the ROM and the direct numerical perturbation（DNP） are consistent. Therefore, ASA and the FSA based on the ROM developed on the 2D/1D transport code KYCORE are verified.

Research on 2D/3D Coupling Method Based on MOC Method

Liang Liang, Liu Zhouyu, Wu Hongchun, Zhang Qian, Zhao Qiang, Zhang Zhijian

2018, 39(S2): 20-24. doi: 10.13832/j.jnpe.2018.S2.0020

Abstract(12) PDF(0)

Abstract:
Recently most of the high-fidelity reactor physics calculation used the characteristics based 2D/1D coupling method as a neutron transport solver. In the classical 2D/1D method, the accuracy of the axial leakage will have direct effect on the final calculation precision. In order to obtain more accurate axial leakage, the 2D/3D coupling method employing the global 3D SN method to get the axial leakage for 2D MOC calculation, meanwhile the 2D MOC calculation supplies the homogenization cross section for 3D SN calculation. To make sure that the 3D SN calculation can take consideration of the space distribution of the flux in each pin cell, a flux correct factor is introduced when computing the outgoing angular flux in 3D SN calculation. In 2D/3D coupling method, iteration between the 2D MOC and 3D SN is implemented to improve the accuracy. In the paper, the detailed theory of this 2D/3D coupling method is introduced, which includes the derivation of the coupling equations and the iterative calculation flow, then the impact of axial leakage to the transport results is analyzed. Based on the theory a 2D/3D coupling code is developed and the results of the 3D C5G7 problems indicates that with less MOC calculation layers the 2D/3D coupling method can obtain accuracy solution and be able to solve small whole-core transport problem.

Correction of Negative Angular Distributions in Multigroup Nuclear Data

Hu Zehua

2018, 39(S2): 25-28. doi: 10.13832/j.jnpe.2018.S2.0025

Abstract(10) PDF(0)

Abstract:
The angular distribution of the secondary neutron is typically represented as a truncated Legendre series expansion. Limited terms of the expansions may result in some non-physical negative regions for highly anisotropic scattering. The positive representation of the scattering angular distribution is produced by the maximum entropy method from the truncated Legendre series. The positive angular distributions are convert to multigroup nuclear data in ACE format available for MCNP. The reliability of the maximum entropy method for correcting the angular distribution is checked by the calculations of criticality benchmark experiment.

Design and Verification of a PWR-Core Pin-by-Pin Fuel Management Calculation Code

Li Yunzhao, Yang Wen, Wang Sicheng, Zhang Bin, Wu Hongchun, Cao Liangzhi

2018, 39(S2): 29-32. doi: 10.13832/j.jnpe.2018.S2.0029

Abstract(10) PDF(0)

Abstract:
PWR-core pin-by-pin fuel management calculation code NECP-Bamboo2.0 employs the Generalized Equivalence Theory to homogenize each pin-cell within the lattice, and employs the Exponential Function Expansion Nodal SP3 method to fulfill the three-dimensional multi-group whole-core neutronics calculation. A whole-core multi-physics parallelization algorithm is designed to carry out the tight coupling between neutronics, thermal hydraulics and depletion calculations. The BEAVRS problem was employed to verify its accuracy. Numerical results demonstrate its good accuracy and flexibility to engineering problems.

Validation of SARAX Code System Using Phenix Control Rod Withdrawal End-of-Life Experiments

Zhou Hang, Zheng Youqi, Hu Yun

2018, 39(S2): 33-37. doi: 10.13832/j.jnpe.2018.S2.0033

Abstract(10) PDF(0)

Abstract:
The Phenix control rod withdrawal end-of-life experiment is the last reactor measurement test before the decommissioning of Phenix rector, a French sodium cooled reactor. In the experiments, the value of control rod worth under low power and the radial power distribution in full power state were measured. Through the modeling and calculation of the experiment using SARAX, a sub-computational program system for the fast reactor developed by Xi’an Jiaotong University, the accuracy of the SARAX program system in the physical calculation of the sodium-cooled MOX fuel fast reactor core can be confirmed. In the calculation, ultra-fine group and point-wise cross section were used for energy spectrum calculation, the super-homogeneous（SPH） factor were used for the assembly homogenization calculation, and the multi-group neutron transport nodal method were used for core calculation. The keff, control rods worth, core reactivity coefficient and power distributions were calculated at four critical states. The calculation results show that the results of SARAX are in good agreement with the experimental values, and the calculation accuracy is better than the traditional fast reactor physical calculation program, which can be used in the nuclear design of sodium-cooled fast reactor loaded with MOX fuel.

Computational Method Study for Effective Self-Shielding Cross-Section of URR in Fast Reactor Code SARAX

Wei Linfang, Wu Hongchun, Zheng Youqi, Du Xianan, Yuan Xianbao

2018, 39(S2): 38-42. doi: 10.13832/j.jnpe.2018.S2.0038

Abstract(12) PDF(0)

Abstract:
In fast reactors, the neutron flux level in Unresolved Resonance Region（URR） of ²³⁸U is relatively high, and the accuracy of cross-section in corresponding energy regions makes direct influence on the fast reactor spectrum and homogenized cross-section calculation. Aiming at effective self-shielding cross-section calculation in URR of ²³⁸U, this paper improves the spectrum calculation code TULIP in SARAX. TULIP uses background cross-section interpolation in URR, but the accuracy of interpolation significantly affects the final results. Considering that the content ratio of ²³⁸U in fast reactor assemblies is large and the background cross-section of ²³⁸U is small, this paper analyzes the requirements of cross-section interpolation for ²³⁸U in URR, and verifies the effectiveness of improving calculation accuracy with denser background cross-section interpolation points by comparing the reaction rate before and after making denser interpolation points of ²³⁸U. Then, the accuracy of SARAX code at high temperatures is improved.

Automated Validation of CGN Nuclear Software Package PCM

Wang Chao, Yang Shuoyan, Peng Sitao, Li Guoren, Ma Yunfan, Chen Jun, Wang Junling, Lu Haoliang

2018, 39(S2): 43-46. doi: 10.13832/j.jnpe.2018.S2.0043

Abstract(10) PDF(0)

Abstract:
The software package PCM, consisted of lattice code PINE, core physics code COCO and flux mapping code MAPLE, has been used in the design of HPR1000. The Validation of PCM used power plant data, experimental data, benchmarks, and similar software’s results. To complete the iterative validation more efficiently, an automated validation platform YAM was developed. The YAM automates the validation process and computational procedure of the software package PCM. The validation shows that the results of PCM match well with the references. The uncertainty of PCM package is less than that assumed in the reactor safety analysis.

Development of On-Line Monitoring Function for SOMAPS

Bi Guangwen, Tang Chuntao, Yang Bo, Fei Jingran, Yang Weiyan, Wang Guozhong

2018, 39(S2): 47-50. doi: 10.13832/j.jnpe.2018.S2.0047

Abstract(14) PDF(0)

Abstract:
The on-line core monitoring system is an important operation supporting system for nuclear reactors of Generation Ⅲnuclear power plants. The SOMPAS（SNERDI On-line Core Monitoring, Prediction and Analysis System）, integrated with advanced three dimensional（3-D） neutronics and thermal hydraulics engine for on-line monitoring core condition, uses plant measurement data especially including in-core neutron detectors to monitor core power distribution and safety margin, which helps to improve the flexibility and to support the optimization of reactor operation. The development and verification of on-line monitoring function of the SOMPAS system are introduced. Numerical simulation and testing verify that SOMPAS monitoring function can provide reliable results, satisfying the expected requirements.

Development and Verification of Resonance Elastic Scattering Kernel Data Processing Module in Nuclear Data Processing Code NECP-Atlas

Xu Jialong, Zu Tiejun, Cao Liangzhi, Wu Hongchun

2018, 39(S2): 51-56. doi: 10.13832/j.jnpe.2018.S2.0051

Abstract(9) PDF(0)

Abstract:
The thermal motion of target nuclei and the resonance elastic scattering have been considered in the generation of multi-group cross sections and scattering matrices. Firstly, the resonance elastic scattering kernel（RESK） formulations for anisotropic scattering up to any Legendre order has been adopted to calculate the exact Doppler broadened energy transfer kernels. A semi-analytical integration method is applied to perform the RESK calculations. Combining with the RESK calculation, a linearization algorithm is proposed to generate the RESK interpolation tables. The RESK data can be interpolated precisely based on the RESK interpolation tables to reduce the calculation burden. Secondly, a neutron slowing-down equation solver is developed based on the RESK instead of the conventional asymptotic scattering kernel. The effect of neutron up-scattering on the neutron energy spectrum can be exactly taken into account by the solver. More precise multi-group cross sections are obtained when more precise energy spectrum are adopted in the group collapsing procedures. All the methods mentioned above have been implemented into the nuclear data processing code called NECP-Atlas. Numerical results show that the proposed methods are capable of producing accurate multi-group cross sections for downstream calculation; Comparedwith the multi-group cross sections and scattering matrices generated by the conventional methods, the fuel Doppler coefficients and eigenvalues calculated by the deterministic codes change greatly when the up-scattering effect is considered.

Delayed γ Dose Calculation Code JMCT-PK

Feng Jingchao, Shi Dunfu, Li Rui, Li Xunzhao, Fu Yuanguang, Zhang Baoyin, Li Gang, Deng Li

2018, 39(S2): 57-61. doi: 10.13832/j.jnpe.2018.S2.0057

Abstract(9) PDF(0)

Abstract:
Based on the JCOGIN（J Combinatorial Geometry Infrastructure） infrastructure and stochastic integration, the point-kernel code JMCT-PK is developed. JMCT-PK supports the CAD modeling with constructive solid geometry. It can obtain the delayed γ source intensity from the calculated result of JMCT（Joined Monte Carlo neutron and photon Transport code）. And JMCT-PK shares the same geometry model with JMCT. This paper introduces the theory of point-kernel method, the programming structure and the verification of the code JMCT-PK. The calculated result of JMCT-PK has a good agreement with the calculated result of VisiPlan, which shows the reliability of the code. The difference shows the importance of build-up factor in the point kernel method.

Effects of Dual-Enrichment Fuel Management on Neutronic Data in Reactor Core

Peng Jinghan

2018, 39(S2): 62-66. doi: 10.13832/j.jnpe.2018.S2.0062

Abstract(15) PDF(0)

Abstract:
This paper discusses the effects of different proportions of fresh fuel assemblies number on neutronic data with SCIENCE V2 package. A fixed loading parttern with 64 fresh fuel assemblies is applied in order to reach a equilibrium state by adjustment of enrichment and Gadolinium rod numbers in fresh fuel assemblies. There are five typical proportion of fresh fuel assemblies with the enrichments of 4.45% and 4.95%, respectively. The effects of different proportion on the generic nuclear data, key neutronic parameters and specific accidental neutronic data are provided. The results show a high flexibility of cycle length. The propotion of fresh fuel assemblies of different enrichements has no effect on the limits of general nuclear data, key neutronic parameters and specific accidental neutronic data.

Study of High-Order Resonance Interference Effect between Adjacent Fuel Rods in Pressurized Water Reactor

Zhang Qian, Zhao Qiang, Liang Liang, Wu Hongchun, Cao Liangzhi

2018, 39(S2): 67-71. doi: 10.13832/j.jnpe.2018.S2.0067

Abstract(18) PDF(0)

Abstract:
In the fuel assmebly of the presurized water reactor, the high-order resonance interference effect between adjacent fuel rods of different fuel types is studied. Based on the method of solving ultra-fine-group slowing-down equation and the embedded resonance calculation method, the specific type of problems is calculated. Numerical results show that the high-order resonance interference effect increases with the reduction of the moderator density and has some influence on the precision of resonance calculation. The maximum relative cross section error reaches 10% in the fuel assembly mixed with uranium dioxide fuel and plutonium uranium oxide fuel. By improving the embedded resonance calculation method, the high-order resonance interference phenomenon can be treated effectively, and the resonance calculation precision of the fuel assembly with complex design can be improved. Results show that the relative error of the resonant section is reduced to less than 3% in the multi-fuel mixing components under various working conditions.

Innovation of Steam Generator Pilot In-Service Inspection Management Strategy

Li Qiuda, Fang Jiang, Gao Fei, Yuan Jianzhong, He Ziang

2018, 39(S2): 72-76. doi: 10.13832/j.jnpe.2018.S2.0072

Abstract(14) PDF(0)

Abstract:
In this paper, through comprehensive technology, combined with peer experience,from the manufacturing, operation, inspection technology to analysis, the strategy of pilot in-service inspection management is put forward. Comparing factory documents, historical data, operation conditions, and defect quantities, the typical representative steam generator（SG）is selected. The reasonable and feasible SG inspection plan is formulated, and has been examined and approved by the regulator. Continuous tracking inspection is carried out for the pilot SG, to provide technical reference for other similar components. According to the inspection results to optimize other SG inspection plan, it has reduced 63000 times of MRPC inspections, and shortened the critical path period of the outage by 60 days. It indirectly creates the economic benefits of ten million yuan for the enterprise, and provides new thoughts and methods for the supervision and management of other in-service inspection components.

Analysis of Flow Streaming and Ex-Core Detector Shadowing Effects Due to Mechanical Shim Operation for AP1000 Nuclear Power Plants

Shi Jianfeng, Gao Mingmin, Yang Qingxiang

2018, 39(S2): 77-81. doi: 10.13832/j.jnpe.2018.S2.0077

Abstract(17) PDF(0)

Abstract:
AP1000 nuclear power plants use Mechanical Shim（MSHIM） operation. At present, few studies have been done on impacts to ex-core operation parameters from MSHIM operation. Control rod insertion and cold leg temperature reduction processes were simulated, and flow streaming and ex-core detector shadowing effects due to MSHIM operation for AP1000 plants were analyzed in this paper. The results show that: M banks can obviously influence the flow streaming and ex-core detector shadowing effects, while AO bank has negligible influence to these effects; cold leg temperature reduction results in obvious ex-core detector shadowing effect which is insensitive to the control rod position. This paper proposed that following control rod positions should be included in related startup tests: 130, 14,-66,-116,-296 and-436 steps. During full power operation, tests at-436 steps could be executed at 130 steps to avoid a low power margin.

Study on Optimum Design of ¹⁴C Source Term in EPR Reactor

Fu Pengtao

2018, 39(S2): 82-86. doi: 10.13832/j.jnpe.2018.S2.0082

Abstract(14) PDF(0)

Abstract:
Production mechanism and calculation method of ¹⁴C in EPR has been introduced in the paper. The normalized emission range of gas phase ¹⁴C is obtained by statistical analysis of large amount of gas phase ¹⁴C emissions from Siemens PWR, and the total amount of ¹⁴C in liquid phase and solid waste is further evaluated. The results show that the annual expected and maximum discharge ranges of ¹⁴C from EPR reactor are respectively 331 GBq/a and 660-700 GBq/a for gaseous discharge, 30 GBq/a and 60GBq/a for liquid discharge, and 64GBq/a and 130GBq/a for solid waste. The calculated gaseous ¹⁴C discharge with 10 ppm dissolved nitrogen in the primary circuits agrees well with the median value in Siemens/KWU PWRs. It reveals that the dissolved nitrogen concentration in the primary circuits is much less than the saturation concentration in volume and control tank covered by nitrogen gas during normal operation. The approaches used and conclusions are also important to the third generation nuclear power plants, including EPR and HPR1000.

Research on Core Design of Pebble Bed Advanced High Temperature Reactor

Wang Lianjie, Sun Wei, Xia Bangyang, Zou Yang, Yan Rui

2018, 39(S2): 87-91. doi: 10.13832/j.jnpe.2018.S2.0087

Abstract(11) PDF(0)

Abstract:
We find that eliminating molten salt and injecting poison into coolant both can be used as the auxiliary system of the second shutdown system in Pebble Bed Advanced High Temperature Reactor（PB-AHTR）. Compared with injecting poison into coolant, the effect of eliminating molten salt on the core is smaller and more beneficial to the engineering realization. Compared to one time charging scheme, batch charging scheme can ensure that the core excess reactivity is small in the whole life, which is more easy to control, but more complex to operate. In one time charging scheme, the second shutdown system gains enough fast shutdown margin by increasing the number of second control rod, not by reducing the stack height of core activity. A PB-AHTR core with burn up life of 100 equivalent full power days is proposed. Both the shutdown margins of the first shutdown system and the second shutdown system meet the design requirements.

One Step Whole Core Calculation of 3D Complex Geometrical Small PWR Based on NECP-X

Cao Lu, Liu Zhouyu, Cao Liangzhi, He Qingming, Wu Hongchun, Han Yu, Bi Guangwen, Tang Chuntao

2018, 39(S2): 92-97. doi: 10.13832/j.jnpe.2018.S2.0092

Abstract(14) PDF(0)

Abstract:
Considering that the conventional typical geometric processing method used in high-fidelity neutronics code NECP-X can hardly handle the complex small reactor, hexagonal fast reactor and so on. CSG（Constructive Solid Geometry） method, which models complicated geometry by the Boolean operation of some simple surfaces, has many advantages, such as geometry flexibility, small memory requirement and easy to be extended. Therefore, the CSG method is applied to NECP-X in this paper, which extends the geometric processing capability. The resonance and transport methods are updated to use the CSG method, and the auto-mesh function is developed to generate meshes with only several parameters. The information of characteristic segments is obtained by the intersection of lines and the constructed geometries. The numerical results show that different reactor cores can be modeled in detail and calculated with the one-step transport method, and the computational results agree well with MC（Monte Carlo） code.

Analysis of Reactivity Perturbation for Monocrystalline Silicon Moving into HFETR

Li Songfa, Zhao Jiaqiang, Xiao Dun, He Chuan, Wang Jiangwen, Zou Quan

2018, 39(S2): 98-102. doi: 10.13832/j.jnpe.2018.S2.0098

Abstract(13) PDF(0)

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When mono-crystalline silicon moving into HFETR core through the irradiation channel, the reactivity disturbance is introduced and the local neutron flux is affected. The mathematical calculation model of HFETR is established through the combination of Monte Carlo N–Particle code（MCNP5） and Monte Carlo N–Particle extended code（MCNPX2.6）, and the availability of the model are verified by the reactor physical start-up experiment. The reactivity disturbance introduced by different quality of mono-crystalline silicon moving into HFETR core from 8^# channel are calculated, and the effects of local neutron flux in neighboring ionization chamber channel when 8 kg mono-crystalline silicon moving into HFETR core from 8# channel are analyzed. The results show that the reactivity disturbance introduced by mono-crystalline silicon moving into core is small and meets the safety requirements of HFETR, however, it will lead to the distortion of neutron flux distribution in adjacent channel to affect neutron detector.

Research on Radioactive Consequences of Severe Accident of China Engineering Test Reactor

Feng Jian, Liu Aihua, Shen Haibo, Qiu Liqing, Zhao Guozheng, Yuan Zhimin, Qing Mingbing

2018, 39(S2): 103-106. doi: 10.13832/j.jnpe.2018.S2.0103

Abstract(10) PDF(0)

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In order to ensure that the radioactive consequences of China Engineering Test Reactor meet the requirements of safety supervision under the severe accident condition and that no out-of-site intervention measures are required, it is necessary to design the discharging scheme of radioactive substances under the severe accident condition. This paper mainly analyzes the leakage rate, the air exhaust volume and the sealed mode, and studies the change rules of radioactive consequences. The results show that the airtight building leakage rate of 2%/d and the discharging scheme of "the emergency ventilation system is closed during the preliminary stage and opened after the 8 h sealed time" should be adopted for China Engineering Test Reactor.

Study on Main Pump Sound Monitoring Technology Based on Matlab

Lai Lisi, Wang Wenlong, Zhang Jiangyun, Zhao Peng, Shi Leigang, Cai Wenchao, Zhang Yun

2018, 39(S2): 107-111. doi: 10.13832/j.jnpe.2018.S2.0107

Abstract(11) PDF(0)

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In order to monitor the operation status of the main pump of the reactor, this paper designs a sound acquisition device for the project by using common tools such as pickup equipment and recording equipment from the point of view of simplicity and effectiveness, and collect the sound of the environment, motor end bearing and pump end bearing of four main pumps in different ways. The frequency spectrum of the collected sound signal is analyzed by MATLAB software, and the frequency and wavelength data are obtained. Then the operation status of the main pump is obtained by waveform comparison and spectrum data analysis. The results show that the sound frequencies of the four main pump bearings are mostly concentrated in the low frequency band, and there are abnormalities in the operation sounds of the 2# main pump and the 4# main pump bearings. The results are consistent with the actual situation.

RELAP5 Model and Test Evaluation of High Flux Engineering Test Reactor

Li Haitao, Zhou Chunlin, Zou Deguang, Zhang Jiangyun, Wang Wenlong, Xiao Dun, Jiang Tinglan

2018, 39(S2): 112-115. doi: 10.13832/j.jnpe.2018.S2.0112

Abstract(19) PDF(0)

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In the safety assessment of High Flux Engineering Test Reactor（HFETR）, the Loss of-Coolant Accident（LOCA） is one of the most important initial events. This paper establishes a numerical calculation model of HFETR reactor based on RELAP5, and simulates the testing case of LOCA. By opening the DN50 valve of primary loop degassing system, the LOCA was simulated. The parameters of reactor inlet and outlet pressures, regulator pressure, and breaking flow rate were tested. Meanwhile, through comparison and analysis of testing data and calculation data, the rationality of the calculation result of RELAP5 program was evaluated. The results show that the RELAP5 calculation result is in good agreement with the experimental results, and the maximum relative error is 7.4%. Those results imply that it is feasible to use the RELAP5 program to simulate the system transients at a low-pressure, pressurized water reactor, such as HFETR.

Study on Response to Full Open Fault of Degassing Flow Control Valve of HFETR

Li Ketian, Li Haitao, Liu Peng, Zou Deguang, Cai Wenchao, Zhao Peng, Lai Lisi

2018, 39(S2): 116-120. doi: 10.13832/j.jnpe.2018.S2.0116

Abstract(16) PDF(0)

Abstract:
This paper builds a primary loop thermal-hydraulic calculation model of High Flux Engineering Test Reactor（HFETR） based on Relap5, and the model is proved to be effective and accurate by an experiment in cold state. By simulating the full open fault of degassing flow control valve（DFCV） happened in running state with this model, the varieties of inlet pressure, outlet pressure and water level of volume compensator are obtained. The results of modeling show that fuel rods of HFETR are confirmed to be well submerged and cooled down, and HFETR remains safe. The reaction time for reactor operators and the reasonable water level of volume compensator are given by analyzing the calculation results, and an advice on DFCV working sate adjustment is proposed.

Cleanness Period Estimation of MJTR Primary Heat Exchanger

Liang Guangyuan, Liu Shuiqing, Xu Taozhong

2018, 39(S2): 121-123. doi: 10.13832/j.jnpe.2018.S2.0121

Abstract(16) PDF(0)

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Primary heat exchanger is the main equipment of MJTR. As the heat transfer capacity is the main factor that affect the safety of the reactor operation, the heat exchanger need to be cleared after some time of operation. In this paper, for the total heat exchange coefficient of MJTR primary heat exchanger, the inlet and outlet water temperature of the primary and secondary cooling system are known, through calculating the logarithmic average temperature margin, heat exchange area and considering the pressure margin of the secondary cooling system, it is shown that when the differential pressure of the second side is greater than or equal to 35.15 kPa, the ability of the primary heat exchanger has descended, and the primary heat exchanger should be cleaned.

Analysis for Core Reactivity of Reduced Enrichment MJTR

Cao Yin, Luo Xin, Wang Hao, Kang Zhanghu, Liu Shuiqing

2018, 39(S2): 124-127. doi: 10.13832/j.jnpe.2018.S2.0124

Abstract(13) PDF(0)

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After the use of low-enrichment fuel in MJTR, the loading of the core has been changed. The core parameters are used to calculate the change rule of ¹³⁵Xe and ¹⁴⁹Sm, and the effect on the core reactivity. Then the curves of core reactivity in nominal operating condition and shut down condition are given, which makes the operator understand the change rule of toxic reactivity and response to the change of working condition in actual operation and ensure the safe operation of the reactor.

Analysis and Study for Change of MJTR Critical Position of Control Rod

Cao Yin, Luo Xin, Liu Shuiqing, Kang Zhanghu, Wang Hao

2018, 39(S2): 128-130. doi: 10.13832/j.jnpe.2018.S2.0128

Abstract(17) PDF(0)

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From 26 th June, 2013 to 14 th June, 2017, dozens of criticality experiments were conducted on MJTR, and the experiment results show that the critical position of control rod are rising. To this phenomenon, some analysis and calculation have been done, and the results show that this phenomenon is a result of several elements such as fission products decay, temperature effect and the reducing of fissile nuclide. This conclusion is of great importance to the safe operation of MJTR in the future.

Design and Safety Analysis of Low-Enriched Core of MJTR

Luo Xin, Cao Yin, Wang Hao, Xu Taozhong, Yang Bin, Deng Caiyu, Kang Zhanghu

2018, 39(S2): 131-134. doi: 10.13832/j.jnpe.2018.S2.0131

Abstract(15) PDF(0)

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After the using of the low-enriched fuel in HFETR, the MJTR can only use the low-enriched fuel. In order to enable MJTR to continue its important role in scientific research, production and technological development, four schemes of low-enriched cores have been designed in this paper, and the analysis of physics, thermo-hydraulics and safety have been conducted. The results of these analyses indicate that, the design of low-enriched cores is able to meet the task requirements, and the safety of the reactor is ensured.

Effect of Different Water Processing Parameters on Water Quality

Yang Song, Tang Yuanhui, Lei Chao, Wei Daming, Chen Qingxu

2018, 39(S2): 135-137. doi: 10.13832/j.jnpe.2018.S2.0135

Abstract(8) PDF(0)

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The reactor requires high quality of the water purifier system, and the pH scale and conductivity is important indicator of water. This study is based on the existing water purifier system, controlling the pH scale by changing the processing parameters. The effect of different processing parameter on water quality is studied, and the processing parameters of the best water quality are analyzed to ensure the water quality of the engineering test reactor.

Failure Analysis on Welded Joint of Valve Nozzle and Branch Pipe in Main Pipe Line of Nuclear Power Plants

Zhao Yongming

2018, 39(S2): 138-141. doi: 10.13832/j.jnpe.2018.S2.0138

Abstract(15) PDF(0)

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The partial penetrating crack in the weld fusion zone of the measurement tube branch pipe and a root valve pipe in the main system secondary loop flow was analyzed in Qinshan Nuclear Power Plant. The macro and scanning electron microscope（SEM） on the end face was observed. Element component analysis was carried out on the different area of the end face. The microstructure and grain size analysis was carried out on the joint section. Joint finite element analysis combined with the pipeline running status came to the conclusion that the crack initiation is caused by the fatigue. The impulse by the change of the fluid flow state causes the valve to vibrate when the fluid flows through the valve. This makes the valve to bear a certain frequency of pull-pull stress. Because there is a large internal pressure, a great stress concentration is produced on the internal wall of the root valve. Fatigue cracks are generated at the root position of the welding seam. After the crack is produced, it is developed into the corrosion fatigue crack under the action of internal fluids.

Analysis and Treatment of Abnormal Vibration of High Pressure Heater to Deaerator Pipeline

Yang Jichen, Song Mingliang, Liang Congcong, Li Zhen, Cao Binbin, Sun Jun

2018, 39(S2): 142-145. doi: 10.13832/j.jnpe.2018.S2.0142

Abstract(11) PDF(0)

Abstract:
There has been a severe problem of hydrophobic pipeline vibration in the pipeline of High-pressure heater to the deaerator since the commissioning of Fujian Fuqing Nuclear Power Plant Unit 2. The unit was forced to decrease the power due to the pipe vibration, which has a serious impact on the stable operation of power plant equipment and the economic benefits of power plant. This paper introduces the vibration phenomena of the pipeline in detail and analyzes the causes of vibration from the aspects of system parameters change, equipment structure and the installation of field equipment. Finally, the abnormal vibration of high-altitude deaerator pipeline can be eliminated by pertinent measures, which can ensure the long-term safe operation of pipeline system. The results can be used for reference in dealing with the vibration of pipes of similar units.

Engineering Application Problems of Degradable Waste Processing Facilities in Tianwan Nuclear Power Plant and Resolutions of the Problems

Liu Hongyu, Li Guanghua, Liu Shenglong

2018, 39(S2): 146-149. doi: 10.13832/j.jnpe.2018.S2.0146

Abstract(12) PDF(0)

Abstract:
Tianwan Nuclear Power Plant is the first nuclear power plant in China to introduce degradable protective articles and its processing facilities. This set of processing facilities was also installed for the first time in the radiation control area of the nuclear island of the nuclear power plant. We have encountered a series of problems in engineering design, installation and commissioning. This paper reviews and analyzes the problems encountered in the engineering practice of the degradable waste processing facilities and the process of solving these problems and also proposes suggestions for the future application of the degradable waste processing facilities facilities.

Study on Thermodynamic Performance for Online Calculation of Nuclear Power Plants

Bao Xudong, Zhao Jindong, Tian Yu

2018, 39(S2): 150-153. doi: 10.13832/j.jnpe.2018.S2.0150

Abstract(12) PDF(0)

Abstract:
Combined with thermal system characteristics and actual operation of nuclear power steam turbine, different methods for calculating the enthalpy of the wet steam are studied. According to the turbine manufacturer design data and unit operation data, based on the method of assumption of wet steam dryness, the calculation model of the operating economy indexes of turbine-generator set is established. Through on-line continuous monitoring of nuclear power plant operation economy indexes, and the analysis of operation performance data of thermal system and equipments, the method of fault diagnosis for thermal system is studied. It can provide technical measures for finding and solving the energy loss problem in the actual operation of the secondary circuit system.

Cause Analysis and Treatment of Pipeline Vibration of Check Valve in Fuqing Nuclear Power Plant

Liang Congcong, Song Mingliang, Zeng Xiaokang

2018, 39(S2): 154-156. doi: 10.13832/j.jnpe.2018.S2.0154

Abstract(16) PDF(0)

Abstract:
The auxiliary steam converter system（STR）of Unit 2 of Fuqing Nuclear Power Plant drains from the drain to the deaerator line during normal power operation. The severe vibration of the pipeline poses potential hidden dangers to the safety, stability and economic operation of the nuclear power plant. According to the analysis of the vibration problem of this pipeline, the theoretical analysis of the amount of hydrophobicity during normal power operation is carried out to find out the root cause of the severe vibration of the analysis pipeline, and corresponding special treatment measures are formulated according to the analysis and the experience of vibration treatment at home and abroad, and finally the vibration is eliminated completely to make the system safe and reliable.

Demonstration of Ameliorate Volumetric Testing in Safe End Welds

Zheng Dexu, Liu Guigang

2018, 39(S2): 157-159. doi: 10.13832/j.jnpe.2018.S2.0157

Abstract(19) PDF(0)

Abstract:
The main pump used in some nuclear plant has safe ends between the main pump and the main pipe. This paper has studied ultrasonic and radiographic testing techniques. RSE-M and ASME standards were analyzed comparatively, and the engineering practice factors were also discussed. It is proved that UT is accurate, continuous and economical for in-service inspection and supervision of the main pump safe ends crack which is the most common and dangerous.

Research on Fuel Management Strategy of Changjiang Nuclear Power Plant under “Large Machine and Small Grid” Operation Environment

Liu Mingquan, Xue Xiang, Wu Danlei

2018, 39(S2): 160-163. doi: 10.13832/j.jnpe.2018.S2.0160

Abstract(9) PDF(0)

Abstract:
The problem that "large machine and small grid" in Hainan Changjiang Nuclear Power Plant is prominent. The load factor is greatly lower than the average level of nuclear power units in China, due to the long-term low power operation, which has great influence on the safety and economy of the unit. It is important for improving the safety and economy to study the extended nuclear fuel cycle. In this paper, the feasibility of the extended nuclear fuel cycle in Changjiang Nuclear Power Plant is studied, and the main safety parameters of the reactor core are calculated. The results show that the long cycle fuel management strategy can enhance the utilization rate of the unit, while meeting the requirements of the safety of the unit.

Study on Boron Saturation and Release Characteristics of Nuclear Grade Resins

Lei Shuixiong, Zou Shiwei, Wang Dong

2018, 39(S2): 164-167. doi: 10.13832/j.jnpe.2018.S2.0164

Abstract(12) PDF(0)

Abstract:
In this paper, the saturation and release properties of nuclear grade lithium mixed bed resins used in the chemical and volume control system（RCV） of nuclear power plant, i.e. the reciprocal transformation relationship between resin hydroxide ion（OH-） and borate ion, are studied, and simulated tests such as the relationship between boron saturation and boron concentration, the influence of different velocity, and the leaching release and re-absorption of saturated resin boron were carried out. The result shows that the boron saturation of ion exchange resin is a dynamic equilibrium, and boron will be absorbed or released with the change of boron concentration in the system; and the change of velocity in a certain range do not cause obvious change to the boron saturation equilibrium time and exchange capacity of resin.

Evaluation of Reactor Pressure Vessel Irradiation Embrittlement of VVER-1000 Nuclear Power Unit

Zhang Meng, Peng Sitong, Wang Chen, Lei Chao

2018, 39(S2): 168-172. doi: 10.13832/j.jnpe.2018.S2.0168

Abstract(17) PDF(0)

Abstract:
Irradiation surveillance and embrittlement evaluation of reactor pressure vessel is an important means to ensure the safe operation during reactor lifetime. This paper introduces the designation and test types of Tianwan NPP Unit 1 and 2（VVER-1000） unit irradiation surveillance sets. Also, the test results, irradiation embrittlement estimation models and lead factor are analyzed and discussed for the 1Л and 2Л sets of Tianwan NPP Unit 1. The irradiation surveillance data shows that the irradiation embrittlement effect of base metal and welds of RPV active area of Tianwan NPP Unit 1 is in the scope of the initial design. The RPV actual embrittlement trend is in good agreement with the estimation model. It is suggested that next irradiation surveillance set be unloaded during the 20 th year of reactor operation.

Application of Discontinuity Factor in the Process of ThreeDimensional Reactor Core Transient Simulation

Duan Xinhui, Wang Bingshu, Ji Li

2018, 39(S2): 173-179. doi: 10.13832/j.jnpe.2018.S2.0173

Abstract(8) PDF(0)

Abstract:
In order to guarantee the static and transient accuracy and real-time performance in the process of the three-dimensional core transient calculation, the high order nodal expansion method is used for the calculation of the fine mesh structure. The results such as average neutron flux and the interface current are considered as the heterogeneous solution. In the homogenization process, the homogeneous group parameters and interface discontinuity factors and boundary albedos are calculated using the flux volume weight method. In the process of transient calculation, according to the control rod changes and the thermal hydraulics parameters feedback, real-time correction is carried out by updating the homogeneous parameters and discontinuity factors. Finally using coarse mesh finite difference method with discontinuity factor correction, a three dimensional core static and transient calculation is realized and verified based on the typical LMW benchmark. Simulation results prove that the accuracy is equivalent to the high order nodal expansion method at both space and time scope. The efficiency is higher than that of the nodal expansion method directly applied in the transient simulation. It meets the requirement of the development for the nuclear power plant full scope simulator.

Verification of Thermal Cooling Capacity for a Spent Fuel Pool Expansion in a 1000 MW PWR Nuclear Power Plant

Zhang Shipeng

2018, 39(S2): 180-184. doi: 10.13832/j.jnpe.2018.S2.0180

Abstract(13) PDF(0)

Abstract:
In view of the capacity expansion project of a 1000 MW PWR NPP spent fuel pool, the thermal cooling capacity of the spent fuel pool after expansion is verified by using the Computational Fluid Dynamics（CFD） and theoretical analysis methods. Under the operating cooling condition of at least one of reactor pool and spent fuel pool cooling and processing system（PTR） in the spent fuel pool, the average water temperature of the spent fuel pool meets the corresponding acceptance criteria. The maximum local water temperature and the maximum temperature of the fuel cladding shell are lower than the local water saturation temperature. On the condition of lost two PTR systems, the time of heating the average water temperature to the boiling temperature and the exposure time of the fuel grid are both calculated, which provides a guidance for the operation intervention.

2018 Vol. 39, No. S2