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检索条件"机构=Science and Technology on Reactor System Design"

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Reflooding characteristics investigation and predicted models improvement for the pressurized water reactor with the tight lattice core design

Reflooding characteristics investigation and predicted model...

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作者： Wu, Dan Yu, Hongxing Yu, Junchong Yu, Jiyang Tsinghua University Beijing 100084 China Science and Technology on Reactor System Design Technology Laboratory Chengdu 610041 China Nuclear Power Institute of China Chengdu 610041 China

The advanced pressurized water reactor with a tight hexagonal lattice core design has its potential to improve uranium utilization. With a smaller pitch-to-diameter ratio (p/d) ranging from 1.05 to 1.20, a major problem of a tight lattice core is the emergency core cooling in a loss-of-coolant accident. The NEPTUN-LWHCR and FLORESTAN reflooding experiments preformed at PSI (Paul Scherrer Institute) and KfK (Kernforschungszentrum Karlsruhe) during the verification period 1985-1990 showed that a LWHCR (light water high conversion reactors) had higher peak cladding temperatures and smaller quench rates than that of a conventional LWR (Rouge and Dreier, 1992;Erbacher and Wiehr, 1988). Furthermore, the existing system codes like RELAP5, CATHARE, TRAC may not be able to simulate the reflooding phase of a LWHCR. In this paper, the characteristics of the reflooding phase of tight lattice cores have been investigated. Then improvements to heat transfer models of precursory cooling have been introduced into RELAP5/MOD3.2. In addition, a new model has been developed to predict the heat release in the vicinity of the quench front. These modifications are assessed using selected tests of NEPTUN-LWHCR and FLORESTAN experiments. Comparisons between the calculated results and measured data indicate that the modified code can predict the peak cladding temperatures and quench temperatures relatively well. While for droplet entrainment prediction, there is still large room for improvement. Further validation is needed for the models proposed in this paper especially for a tight lattice core with a p/d ratio less than 1.10, and more efforts are required to study the droplet distribution and droplet size downstream of the quench front for higher reflooding rate tests (larger than 4.5 cm/s). © 2012 Elsevier B.V.

关键词： Loss of coolant accidents

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A least-squares method based on coupling coefficients for core power distribution monitoring in pressurized water reactors

A least-squares method based on coupling coefficients for co...

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2013 Winter Meeting on Transactions and Embedded Topical Meetings: Risk Management for Complex Socio-Technical systems, 2nd ANS SMR 2013 Conference, Nuclear Nonproliferation - 1st Fission to the Future

作者： Peng, Xingjie Wang, Kan Li, Qing Department of Engineering Physics Tsinghua University Beijing 100084 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu 610041 China

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Multiple choices of reactor core nuclear design for ACP100's application in different scenarios

Multiple choices of reactor core nuclear design for ACP100's...

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2020 International Conference on Physics of reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020

作者： Wang, Liangzi Ju, Haitao Li, Qing Qin, Dong Wang, Lianjie Yu, Yingrui Ning, Zhonghao Wang, Chenlin Guo, Rui Wang, Shuai Zhang, Bin Xiang, Hongzhi Lou, Lei Sun, Wei Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China NO. 328 Section 1 Changshun Ave Shuangliu District Sichuan Chengdu China

ISBN: (纸本)9781713827245

ACP100 NPP designed by CNNC (China National Nuclear Corporation) is a 125MWe, multi-purpose small modular reactor based on pressurized water reactor technology;it adopts the integrated reactor technology. Different application scenarios bring up different design requirements: some require high compactness, but others care more about a longer cycle length, and some may require a fully mature and conservative design;thus, multiple design choices need to be proposed. Also, the same and most important thing cared by all users is that, the design needs to be validated to satisfy the current nuclear safety standards, and lower cost would be always preferred. Core nuclear design is a key part of the whole NPP design. Basically, nuclear design target of ACP100 is to achieve a reasonable good balance during longer cycle length, larger discharge exposure for fuel assemblies, and maximally using the mature technologies, and of course, with sufficient reactivity control ability for safety assurance. Aiming at satisfying all these different needs maximally, a strategy of supplying multiple nuclear design choices is proposed for ACP100: choice 1. Boron-free plan, this is a compact design with no need for chemistry and volume system, no need for da ily boron adjustment and relative waste storage;choice 2. Boron and rod co-controlled plan, this is similar with large commercial PWRs, with a lower power peak factor and suitable for broad location sites. Both choices load 57 units of the same type fuel assemblies CF3S (with height reduced from CF3 fuel assemblies) per cycle, and both adopt partial reload and shuffle fuel management strategy to achieve larger discharge exposure. Gd is loaded in the fuel rods in both choices to help control reactivity. Choice 1 loads much more control rod clusters than choice 2, and of course, reactivity adjustment and compensation during operation is totally different between them. Using suitable and reliable software to simulate t he core, through lar

关键词： Boron

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RMC/CTF multiphysics solutions to VERA benchmark problem 9

RMC/CTF multiphysics solutions to VERA benchmark problem 9

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2018 Transactions of the American Nuclear Society, ANS 2018

作者： Ma, Yugao Liu, Shichang Luo, Zhen Wang, Kan Yu, Ganglin Huang, Shanfang Department of Engineering Physics Tsinghua University Beijing100084 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

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Methods and Verification of 3D direct reactor calculation code KYCORE

Methods and Verification of 3D direct reactor calculation co...

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2018 International Conference on Physics of reactors: reactor Physics Paving the Way Towards More Efficient systems, PHYSOR 2018

作者： Tang, Xiao Li, Qing Chai, Xiaoming Tu, Xiaolan Peng, Xingjie Wang, Kan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Department of Engineering Physics Tsinghua University Beijing 100084 China

ISBN: (纸本)9781713808510

KYCORE is a 3D direct reactor calculation code capable of performing reactor hetero geneous calculation without cross section homogenization and group condense. It is developed in Nuclear Power Institute of China, involving radial characteristics method (MOC) and axial discrete ordinate method (SN) coupled 3D neutron transport and subgroup method with fixed source problem for resonance treatment and depletion, aiming at providing mini core design and commercial reactor benchmark solution. This article briefly introduces optimized MOC/SN coupling, adjusted coarse mesh finite difference (CMFD) and resonance treatment. The accuracy of the code is verified with 3D VERA Problem 3 benchmark and proves to be credible. © 2018 International Conference on Physics of reactors, PHYSOR 2018: reactor Physics Paving the Way Towards More Efficient systems. All rights reserved.

关键词： Codes (symbols)

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Decoupling and Coupling Simulation Analysis in Small Nuclear Power Plant 23rd

Decoupling and Coupling Simulation Analysis in Small Nuclear...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Tian, Peiyu Li, Yi Xing, Tianyang Liang, Tiebo Wang, Changshuo Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China School of Energy and Environment Southeast University Jiangsu Nanjing China

ISBN: (纸本)9789819910229

As the development of computer science technology and the requirement of thorough research, more researchers are setting their eyes on coupling method because of the tight coupling in NPP (nuclear power plant) system. However there were few researches studying the difference between decoupling and coupling methods and the importance of coupling method. This research respectively establishes the primary and secondary loop decoupling models and the two loops coupling model based on the small NPP by APROS. Then the differences between the decoupling and coupling models is studied under the steady state and dynamic state which contains the ramp load variation and load shedding. The results show that there are small differences between these models in the main parameter values under the steady state. But the differences between decoupling models and coupling model are large. Therefore the NPP system needs be modeled by coupling method as to study its dynamic characteristic. © 2023, The Author(s).

关键词： Nuclear power plants

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RMC/ANSYS Multi-physics coupling solutions for heat pipe cooled reactors analyses

RMC/ANSYS Multi-physics coupling solutions for heat pipe coo...

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2020 International Conference on Physics of reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020

作者： Ma, Yugao Liu, Minyun Chen, Erhui Xie, Biheng Chai, Xiaoming Huang, Shanfang Wang, Kan Yu, Hongxing Department of Engineering Physics Tsinghua University Beijing100084 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9781713827245

The heat pipe cooled reactor is a solid-state reactor using heat pipes to passively transfer heat generated from the reactor, which is a potential and near-term space nuclear power system. This paper introduces the coupling scheme between the continuous energy reactor Monte Carlo (RMC) code and the finite element method commercial software ANSYS. Monte Carlo method has the advantages of flexible geometry modeling and continuous-energy nuclear cross sections. ANSYS Parametric design Language (APDL) is used to determine the detailed temperature distributions and geometric deformation. The on-the-fly temperature treatment of cross sections was adopted in RMC code to solve the memory problems and to speed up simulations. This paper proposed a geometric updating strategy and reactivity feedback methods for the geometric deformation of the solid-state core. The neutronic and thermal-mechanical coupling platform is developed to analyze and further to optimize the heat pipe cooled reactor design. The present coupling codes analyze a 2D central cross-section model for MEGAPOWER heat pipe cooled reactor. The thermal-mechanical feedback reveals that the solid-state reactor has a negative reactivity feedback (~1.5 pcm/K) while it has a deterioration in heat transfer due to the expansion. © The Authors, published by EDP sciences. This is an open access article distributed under the terms of the Creative Commons Attribution License 4.0 (http://***/licenses/by/4.0/).

关键词： Finite element method

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Radiolytic gas bubbles' model and its application to AHR's transient neutronics analysis

Radiolytic gas bubbles' model and its application to AHR's t...

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2017 Transactions of the American Nuclear Society, ANS 2017

作者： Liangzi, Wang Dong, Yao Yingrui, Yu Shuai, Wang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China First-section of Chang Shun Avenue No. 328 ChengDu SiChuan610213 China

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Development of a sensitivity analysis code for 3-D C5G7 benchmarks based on the pertubation theary

Development of a sensitivity analysis code for 3-D C5G7 benc...

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2018 International Conference on Physics of reactors: reactor Physics Paving the Way Towards More Efficient systems, PHYSOR 2018

作者： Wu, Qu Peng, Xingjie Yu, Yingrui Li, Qing Wang, Kan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Department of Engineering Physics Tsinghua University Beijing100084 China

ISBN: (纸本)9781713808510

Sensitivity and Uncertainty (S&U) analysis plays a major role in reactor physics. Uncertainty in reactor engineering comes from model error, numerical error and input parameter error. Nuclear data are regarded as crucial input parameters and proven to be a major uncertainty source. Therefore, effective evaluation of nuclear data can promote the accuracy and fidelity of nuclear system simulation. In deterministic codes, however, most studies about S&U analysis are based on the traditional two-step method, i.e. assembly calculation and core simulation. With the development of advanced 2-D/l-D transport codes, the corresponding S&U analysis model is required. A new S&U analysis code, namely KYADJ, achieved the sensitivity analysis of eigenvalue with respect to material macroscopic cross sections based on classical perturbation theory. Forward and adjoint flux were solved by the coupling of 2-D Method of Characteristics (MOC) and 1-D discrete ordinates method (Sn). Besides, the Direct Numerical Perturbation (DNP) method in KYADJ and the Iterated Fission Probability (UP) method in reactor Monte Carlo code (RMC) were adopted to calculate sensitivity coefficients for comparison. A 3x3 pin-cell benchmark and 3-D OECD/NEA C5G7 benchmarks were applied to verify the ability of sensitivity analysis in KYADJ. Results showed that a good agreement among three approaches was achieved. KYADJ can accurately calculate eigenvalue sensitivity coefficients with respect to input nuclear data. © 2018 International Conference on Physics of reactors, PHYSOR 2018: reactor Physics Paving the Way Towards More Efficient systems. All rights reserved.

关键词： Sensitivity analysis

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STARTUP ANALYSES OF A HELIUM-XENON GAS-COOLED reactor 30

STARTUP ANALYSES OF A HELIUM-XENON GAS-COOLED REACTOR

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30th International Conference on Nuclear Engineering, ICONE 2023

作者： Ma, Yugao Bao, Hui Lu, Jiahao Ding, Shuhua Zhang, Meng Yang, Xiaolong Deng, Jian Zhang, Muhao Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China College of Nuclear Technology and Automation Engineering Chengdu University of Technology Chengdu610059 China

ISBN: (纸本)9784888982566

The Helium-Xenon (He-Xe) gas-cooled reactor uses a He-Xe gas mixture to remove the heat from the reactor core. It has the characteristics of a long life-cycle, simple operation, and high power-mass ratio, which has an excellent application prospect in the field of Space reactors Power systems. The high operation temperature distinguishes the He-Xe gas-cooled reactor from the traditional reactor in reactivity feedback characteristics. Except for the Doppler effect of the fuel and monolith, the thermal expansion also introduces reactivity feedback during reactor operation. In this work, reactivity feedback characteristics, steady-state characteristics, and transient characteristics of a hundred-kilowatt-level He-Xe gas-cooled reactor (Submersion-Subcritical Safe Space reactor, S4) were analyzed. The Doppler feedback characteristics and thermal expansion feedback coefficients of fuel, matrix, and reflector from the cold to the hot state of the S4 reactor core were calculated based on the reactor Monte Carlo code (RMC). The results show that the S4 reactor core is mainly affected by the fuel Doppler effect and the matrix thermal expansion effect. Furthermore, the temperature distributions of fuel and He-Xe gas of the S4 reactor were calculated using the He-Xe reactor transient analysis code (HXRTRAN). The transient characteristics of the S4 reactor core when different reactivity was introduced were analyzed. This study may provide a reference for the operation state of the He-Xe gas-cooled reactor. © 2023 by JSME.

关键词： Helium

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