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[Online] Nuclear Power

Guest Editors-in-Chief 
Ye, Qizhen, China National Nuclear Corporation, China
Sehgal, Bal Raj, Royal Institute of Technology, Sweden
 
Members
Ashurko, Yury, Institute for Physics and Power Engineering Named after
A.I.Leypunsky(IPPE), Russia
Barnert, Heiko, Forschungszentrum Jülich, Germany
Guidez, Joel, French Alternative Energies and Atomic Energy Commission, France
Kadak, Andrew, Kadak Associates, Inc., USA
Lohnert, Guenter, University of Stuttgart, Germany
Ma, Weimin, China Nuclear Power Engineering Co., Ltd., China
Machenaud, Herve, Electricite De France (EDF), France
Smith, Craig F., Naval Postgraduate School, USA
Su, Gang, China Nuclear Power Engineering Co., Ltd., China
Wu, Yican, Institute of Nuclear Energy Safety Technology, CAS, China
Xu, Mi, China Institute of Atomic Energy, China
Xu, Yuanhui, Tsinghua University, China
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HPR1000: Advanced Pressurized Water Reactor with Active and Passive Safety
Ji Xing, Daiyong Song, Yuxiang Wu
Engineering    2016, 2 (1): 79-87.   https://doi.org/10.1016/J.ENG.2016.01.017
Abstract   HTML   PDF (2889KB)

HPR1000 is an advanced nuclear power plant (NPP) with the significant feature of an active and passive safety design philosophy, developed by the China National Nuclear Corporation. On one hand, it is an evolutionary design based on proven technology of the existing pressurized water reactor NPP; on the other hand, it incorporates advanced design features including a 177-fuel-assembly core loaded with CF3 fuel assemblies, active and passive safety systems, comprehensive severe accident prevention and mitigation measures, enhanced protection against external events, and improved emergency response capability. Extensive verification experiments and tests have been performed for critical innovative improvements on passive systems, the reactor core, and the main equipment. The design of HPR1000 fulfills the international utility requirements for advanced light water reactors and the latest nuclear safety requirements, and addresses the safety issues relevant to the Fukushima accident. Along with its outstanding safety and economy, HPR1000 provides an excellent and practicable solution for both domestic and international nuclear power markets.

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The Traveling Wave Reactor: Design and Development
John Gilleland, Robert Petroski, Kevan Weaver
Engineering    2016, 2 (1): 88-96.   https://doi.org/10.1016/J.ENG.2016.01.024
Abstract   HTML   PDF (2254KB)

The traveling wave reactor (TWR) is a once-through reactor that uses in situ breeding to greatly reduce the need for enrichment and reprocessing. Breeding converts incoming subcritical reload fuel into new critical fuel, allowing a breed-burn wave to propagate. The concept works on the basis that breed-burn waves and the fuel move relative to one another. Thus either the fuel or the waves may move relative to the stationary observer. The most practical embodiments of the TWR involve moving the fuel while keeping the nuclear reactions in one place−sometimes referred to as the standing wave reactor (SWR). TWRs can operate with uranium reload fuels including totally depleted uranium, natural uranium, and low-enriched fuel (e.g., 5.5% 235U and below), which ordinarily would not be critical in a fast spectrum. Spent light water reactor (LWR) fuel may also serve as TWR reload fuel. In each of these cases, very efficient fuel usage and significant reduction of waste volumes are achieved without the need for reprocessing. The ultimate advantages of the TWR are realized when the reload fuel is depleted uranium, where after the startup period, no enrichment facilities are needed to sustain the first reactor and a chain of successor reactors. TerraPower’s conceptual and engineering design and associated technology development activities have been underway since late 2006, with over 50 institutions working in a highly coordinated effort to place the first unit in operation by 2026. This paper summarizes the TWR technology: its development program, its progress, and an analysis of its social and economic benefits.

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The General Design and Technology Innovations of CAP1400
Mingguang Zheng, Jinquan Yan, Shentu Jun, Lin Tian, Xujia Wang, Zhongming Qiu
Engineering    2016, 2 (1): 97-102.   https://doi.org/10.1016/J.ENG.2016.01.018
Abstract   HTML   PDF (2559KB)

The pressurized water reactor CAP1400 is one of the sixteen National Science and Technology Major Projects. Developed from China’s nuclear R&D system and manufacturing capability, as well as AP1000 technology introduction and assimilation, CAP1400 is an advanced large passive nuclear power plant with independent intellectual property rights. By discussing the top design principle, main performance objectives, general parameters, safety design, and important improvements in safety, economy, and other advanced features, this paper reveals the technology innovation and competitiveness of CAP1400 as an internationally promising Gen-III PWR model. Moreover, the R&D of CAP1400 has greatly promoted China’s domestic nuclear power industry from the Gen-II to the Gen-III level.

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In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs
Weimin Ma, Yidan Yuan, Bal Raj Sehgal
Engineering    2016, 2 (1): 103-111.   https://doi.org/10.1016/J.ENG.2016.01.019
Abstract   HTML   PDF (1699KB)

A historical review of in-vessel melt retention (IVR) is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs). The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential phenomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel (RPV). For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contribute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.

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The Shandong Shidao Bay 200 MWe High-Temperature Gas-Cooled Reactor Pebble-Bed Module (HTR-PM) Demonstration Power Plant: An Engineering and Technological Innovation
Zuoyi Zhang, Yujie Dong, Fu Li, Zhengming Zhang, Haitao Wang, Xiaojin Huang, Hong Li, Bing Liu, Xinxin Wu, Hong Wang, Xingzhong Diao, Haiquan Zhang, Jinhua Wang
Engineering    2016, 2 (1): 112-118.   https://doi.org/10.1016/J.ENG.2016.01.020
Abstract   HTML   PDF (1901KB)

After the first concrete was poured on December 9, 2012 at the Shidao Bay site in Rongcheng, Shandong Province, China, the construction of the reactor building for the world’s first high-temperature gas-cooled reactor pebble-bed module (HTR-PM) demonstration power plant was completed in June, 2015. Installation of the main equipment then began, and the power plant is currently progressing well toward connecting to the grid at the end of 2017. The thermal power of a single HTR-PM reactor module is 250 MWth, the helium temperatures at the reactor core inlet/outlet are 250/750 °C, and a steam of 13.25 MPa/567 °C is produced at the steam generator outlet. Two HTR-PM reactor modules are connected to a steam turbine to form a 210 MWe nuclear power plant. Due to China’s industrial capability, we were able to overcome great difficulties, manufacture first-of-a-kind equipment, and realize series major technological innovations. We have achieved successful results in many aspects, including planning and implementing R&D, establishing an industrial partnership, manufacturing equipment, fuel production, licensing, site preparation, and balancing safety and economics; these obtained experiences may also be referenced by the global nuclear community.

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The Status of the US High-Temperature Gas Reactors
Andrew C. Kadak
Engineering    2016, 2 (1): 119-123.   https://doi.org/10.1016/J.ENG.2016.01.026
Abstract   HTML   PDF (895KB)

In 2005, the US passed the Energy Policy Act of 2005 mandating the construction and operation of a high-temperature gas reactor (HTGR) by 2021. This law was passed after a multiyear study by national experts on what future nuclear technologies should be developed. As a result of the Act, the US Congress chose to develop the so-called Next-Generation Nuclear Plant, which was to be an HTGR designed to produce process heat for hydrogen production. Despite high hopes and expectations, the current status is that high temperature reactors have been relegated to completing research programs on advanced fuels, graphite and materials with no plans to build a demonstration plant as required by the US Congress in 2005. There are many reasons behind this diminution of HTGR development, including but not limited to insufficient government funding requirements for research, unrealistically high temperature requirements for the reactor, the delay in the need for a “hydrogen” economy, competition from light water small modular light water reactors, little utility interest in new technologies, very low natural gas prices in the US, and a challenging licensing process in the US for non-water reactors.

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Design and R&D Progress of China Lead-Based Reactor for ADS Research Facility
Yican Wu
Engineering    2016, 2 (1): 124-131.   https://doi.org/10.1016/J.ENG.2016.01.023
Abstract   HTML   PDF (2944KB)

In 2011, the Chinese Academy of Sciences launched an engineering project to develop an accelerator-driven subcritical system (ADS) for nuclear waste transmutation. The China Lead-based Reactor (CLEAR), proposed by the Institute of Nuclear Energy Safety Technology, was selected as the reference reactor for ADS development, as well as for the technology development of the Generation IV lead-cooled fast reactor. The conceptual design of CLEAR-I with 10 MW thermal power has been completed. KYLIN series lead-bismuth eutectic experimental loops have been constructed to investigate the technologies of the coolant, key components, structural materials, fuel assembly, operation, and control. In order to validate and test the key components and integrated operating technology of the lead-based reactor, the lead alloy-cooled non-nuclear reactor CLEAR-S, the lead-based zero-power nuclear reactor CLEAR-0, and the lead-based virtual reactor CLEAR-V are under realization.

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